National Library of Energy BETA

Sample records for heu highly enriched

  1. German Pebble Bed Research Reactor Highly Enriched Uranium (HEU) Fuel

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    German Pebble Bed Research Reactor Highly Enriched Uranium (HEU) Fuel Environmental Assessment Maxcine Maxted, DOE-SR Used Nuclear Fuel Program Manager June 24, 2014 Public Scoping Meeting

  2. German Pebble Bed Research Reactor Highly Enriched Uranium (HEU...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Potential Acceptance and Disposition of German Pebble Bed Research Reactor Highly Enriched Uranium (HEU) Fuel Environmental Assessment Maxcine Maxted, DOE-SR Used Nuclear Fuel...

  3. Realities of verifying the absence of highly enriched uranium (HEU) in gas centrifuge enrichment plants

    SciTech Connect (OSTI)

    Swindle, D.W.

    1990-03-01

    Over a two and one-half year period beginning in 1981, representatives of six countries (United States, United Kingdom, Federal Republic of Germany, Australia, The Netherlands, and Japan) and the inspectorate organizations of the International Atomic Energy Agency and EURATOM developed and agreed to a technically sound approach for verifying the absence of highly enriched uranium (HEU) in gas centrifuge enrichment plants. This effort, known as the Hexapartite Safeguards Project (HSP), led to the first international concensus on techniques and requirements for effective verification of the absence of weapons-grade nuclear materials production. Since that agreement, research and development has continued on the radiation detection technology-based technique that technically confirms the HSP goal is achievable. However, the realities of achieving the HSP goal of effective technical verification have not yet been fully attained. Issues such as design and operating conditions unique to each gas centrifuge plant, concern about the potential for sensitive technology disclosures, and on-site support requirements have hindered full implementation and operator support of the HSP agreement. In future arms control treaties that may limit or monitor fissile material production, the negotiators must recognize and account for the realities and practicalities in verifying the absence of HEU production. This paper will describe the experiences and realities of trying to achieve the goal of developing and implementing an effective approach for verifying the absence of HEU production. 3 figs.

  4. heu

    National Nuclear Security Administration (NNSA)

    plans to reduce the proliferation threat from stockpiles of surplus highly enriched uranium (HEU) by down-blending, or converting, it to low enriched uranium (LEU). In this...

  5. heu

    National Nuclear Security Administration (NNSA)

    In 1996, the Department of Energy (DOE) announced plans to reduce the proliferation threat from stockpiles of surplus highly enriched uranium (HEU) by down-blending,...

  6. Active Interrogation Observables for Enrichment Determination of DU Shielded HEU Metal Assemblies with Limited Geometrical Information

    SciTech Connect (OSTI)

    Pena, Kirsten E; McConchie, Seth M; Crye, Jason Michael; Mihalczo, John T

    2011-01-01

    Determining the enrichment of highly enriched uranium (HEU) metal assemblies shielded by depleted uranium (DU) proves a unique challenge to currently employed measurement techniques. Efforts to match time-correlated neutron distributions obtained through active interrogation to Monte Carlo simulations of the assemblies have shown promising results, given that the exact geometries of both the HEU metal assemblies and DU shields are known from imaging and fission site mapping. In certain situations, however, it is desirable to obtain enrichment with limited or no geometrical information of the assemblies being measured. This paper explores the possibility that the utilization of observables in the interrogation of assemblies by time-tagged D-T neutrons, including time-correlated distribution of neutrons and gammas using liquid scintillators operating on the fission chain time scale, can lead to enrichment determination without a complete set of geometrical information.

  7. Unattended Environmental Sampling and Laser-based Enrichment Assay for Detection of Undeclared HEU Production in Enrichment Plants

    SciTech Connect (OSTI)

    Anheier, Norman C.; Bushaw, Bruce A.

    2010-04-15

    Nuclear power is enjoying rapid growth as government energy policies and public demand shift toward carbon neutral energy production. Accompanying the growth in nuclear power is the requirement for increased nuclear fuel production, including a significant expansion in uranium enrichment capacity. Essential to the success of the nuclear energy renaissance is the development and implementation of sustainable, proliferation-resistant nuclear power generation. Unauthorized production of highly enriched uranium (HEU) remains the primary proliferation concern for modern gaseous centrifuge enrichment plants (GCEPs). While to date there has been no indication of declared, safeguarded GCEPs producing HEU, the massive separative work unit (SWU) processing power of modern GCEPs presents a significant latent risk of nuclear breakout and suggests the need for more timely detection of potential facility misuse. The Pacific Northwest National Laboratory is developing an unattended safeguards instrument, combining continuous aerosol particulate collection with uranium isotope assay, to provide timely HEU detection within a GCEP. This approach is based on laser vaporization of aerosol particulates, followed by laser spectroscopy to characterize the uranium enrichment level. We demonstrate enrichment assay, with relative isotope abundance uncertainty <5%, on individual micron-sized particles that are trace components within a mixture ‘background’ particles

  8. Portable NDA Equipment for Enrichment Measurements in the HEU Transparency Program

    SciTech Connect (OSTI)

    Decman, D J; Bandong, B B; Wong, J L; Valentine, J D; Luke, S J

    2008-06-02

    The Highly Enriched Uranium (HEU) Transparency Program has used portable nondestructive assay (NDA) equipment to measure the {sup 235}U enrichment of material subject to the transparency agreement since 1997. The equipment is based on the 'enrichment meter' method and uses low-resolution sodium iodide (NaI(Tl)) detectors. Although systems using high-purity germanium (HPGe) detectors can produce more accurate results we have found that the results with NaI(Tl) detectors are quite adequate for the requirements of the transparency agreement. This paper will describe the details of the equipment's operation, calibration, testing, and deployment in Russia. We will also provide a comparison of the units originally deployed in 1997 with the upgraded systems that were deployed in 2003.

  9. Detection of illicit HEU production in gaseous centrifuge enrichment plants using neutron counting techniques on product cylinders

    SciTech Connect (OSTI)

    Freeman, Corey R; Geist, William H

    2010-01-01

    Innovative and novel safeguards approaches are needed for nuclear energy to meet global energy needs without the threat of nuclear weapons proliferation. Part of these efforts will include creating verification techniques that can monitor uranium enrichment facilities for illicit production of highly-enriched uranium (HEU). Passive nondestructive assay (NDA) techniques will be critical in preventing illicit HEU production because NDA offers the possibility of continuous and unattended monitoring capabilities with limited impact on facility operations. Gaseous centrifuge enrichment plants (GCEP) are commonly used to produce low-enriched uranium (LEU) for reactor fuel. In a GCEP, gaseous UF{sub 6} spins at high velocities in centrifuges to separate the molecules containing {sup 238}U from those containing the lighter {sup 235}U. Unfortunately, the process for creating LEU is inherently the same as HEU, creating a proliferation concern. Insuring that GCEPs are producing declared enrichments poses many difficult challenges. In a GCEP, large cascade halls operating thousands of centrifuges work together to enrich the uranium which makes effective monitoring of the cascade hall economically prohibitive and invasive to plant operations. However, the enriched uranium exiting the cascade hall fills product cylinders where the UF{sub 6} gas sublimes and condenses for easier storage and transportation. These product cylinders hold large quantities of enriched uranium, offering a strong signal for NDA measurement. Neutrons have a large penetrability through materials making their use advantageous compared to gamma techniques where the signal is easily attenuated. One proposed technique for detecting HEU production in a GCEP is using neutron coincidence counting at the product cylinder take off stations. This paper discusses findings from Monte Carlo N-Particle eXtended (MCNPX) code simulations that examine the feasibility of such a detector.

  10. Unattended Monitoring of HEU Production in Gaseous Centrifuge Enrichment Plants using Automated Aerosol Collection and Laser-based Enrichment Assay

    SciTech Connect (OSTI)

    Anheier, Norman C.; Bushaw, Bruce A.

    2010-08-11

    Nuclear power is enjoying rapid growth as government energy policies and public demand shift toward low carbon energy production. Pivotal to the global nuclear power renaissance is the development and deployment of robust safeguards instrumentation that allows the limited resources of the IAEA to keep pace with the expansion of the nuclear fuel cycle. Undeclared production of highly enriched uranium (HEU) remains a primary proliferation concern for modern gaseous centrifuge enrichment plants (GCEPs), due to their massive separative work unit (SWU) processing power and comparably short cascade equilibrium timescale. The Pacific Northwest National Laboratory is developing an unattended safeguards instrument, combining continuous aerosol particulate collection with uranium isotope assay, to provide timely detection of HEU production within a GCEP. This approach is based on laser vaporization of aerosol particulates, followed by laser spectroscopy to characterize the uranium enrichment level. Our prior investigation demonstrated single-shot detection sensitivity approaching the femtogram range and relative isotope ratio uncertainty better than 10% using gadolinium as a surrogate for uranium. In this paper we present measurement results on standard samples containing traces of depleted, natural, and low enriched uranium, as well as measurements on aerodynamic size uranium particles mixed in background materials (e.g., dust, minerals, soils). Improvements and optimizations in the detection electronics, signal timing, calibration, and laser alignment have lead to significant improvements in detection sensitivity and enrichment accuracy, contributing to an overall reduction in the false alarm probability. The sample substrate media was also found to play a significant role in facilitating laser-induced vaporization and the production of energetic plasma conditions, resulting in ablation optimization and further improvements in the isotope abundance sensitivity.

  11. Highly Enriched Uranium Transparency Program | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    Administration Highly Enriched Uranium Transparency Program November 13, 2013 The U.S. National Nuclear Security Administration's (NNSA) Highly Enriched Uranium (HEU) Transparency Program reduces nuclear risk by monitoring the conversion of 500 metric tons (MT) of Russian HEU, enough material for 20,000 nuclear weapons, into low enriched uranium (LEU). This LEU is put into peaceful use in the United States, generating nearly 10% of all U.S. electrical power. The HEU Purchase Agreement:

  12. Belgium Highly Enriched Uranium and Plutonium Removals | National Nuclear

    National Nuclear Security Administration (NNSA)

    Security Administration Highly Enriched Uranium and Plutonium Removals March 24, 2014 Belgium has been a global leader in nonproliferation, working with the United States since 2006 to minimize highly enriched uranium (HEU) and plutonium inventories in Belgium through the return of a significant amount of HEU and plutonium to the United States. At the 2014 Nuclear Security Summit, the United States and Belgium announced the successful removal of all excess fresh HEU and plutonium from

  13. Microsoft PowerPoint - 10_BILL_WANDERER_REVISED HEU Agreement NMMSS Briefing_30 April 2014.pptx

    National Nuclear Security Administration (NNSA)

    -Russia Highly Enriched Uranium(HEU) Purchase Agreement Update Bill Wanderer Department of Energy/National Nuclear Security Administration Office of Nuclear Verification HEU Transparency Program The HEU Purchase Agreement 1993 HEU Purchase Agreement - United States purchased low enriched uranium (LEU) derived from 500 metric tons (MT) Russian weapons- origin HEU - Commercial implementation by two Executive Agents * U.S. Executive Agent was the United States Enrichment Corporation (USEC) *

  14. Isotope Enrichment Detection by Laser Ablation - Laser Absorption Spectrometry: Automated Environmental Sampling and Laser-Based Analysis for HEU Detection

    SciTech Connect (OSTI)

    Anheier, Norman C.; Bushaw, Bruce A.

    2010-01-01

    The global expansion of nuclear power, and consequently the uranium enrichment industry, requires the development of new safeguards technology to mitigate proliferation risks. Current enrichment monitoring instruments exist that provide only yes/no detection of highly enriched uranium (HEU) production. More accurate accountancy measurements are typically restricted to gamma-ray and weight measurements taken in cylinder storage yards. Analysis of environmental and cylinder content samples have much higher effectiveness, but this approach requires onsite sampling, shipping, and time-consuming laboratory analysis and reporting. Given that large modern gaseous centrifuge enrichment plants (GCEPs) can quickly produce a significant quantity (SQ ) of HEU, these limitations in verification suggest the need for more timely detection of potential facility misuse. The Pacific Northwest National Laboratory (PNNL) is developing an unattended safeguards instrument concept, combining continuous aerosol particulate collection with uranium isotope assay, to provide timely analysis of enrichment levels within low enriched uranium facilities. This approach is based on laser vaporization of aerosol particulate samples, followed by wavelength tuned laser diode spectroscopy to characterize the uranium isotopic ratio through subtle differences in atomic absorption wavelengths. Environmental sampling (ES) media from an integrated aerosol collector is introduced into a small, reduced pressure chamber, where a focused pulsed laser vaporizes material from a 10 to 20-m diameter spot of the surface of the sampling media. The plume of ejected material begins as high-temperature plasma that yields ions and atoms, as well as molecules and molecular ions. We concentrate on the plume of atomic vapor that remains after the plasma has expanded and then cooled by the surrounding cover gas. Tunable diode lasers are directed through this plume and each isotope is detected by monitoring absorbance signals on a shot-to-shot basis. The media is translated by a micron resolution scanning system, allowing the isotope analysis to cover the entire sample surface. We also report, to the best of our knowledge, the first demonstration of laser-based isotopic measurements on individual micron-sized particles that are minor target components in a much larger heterogeneous mix of background particles. This composition is consistent with swipe and environmental aerosol samples typically collected for safeguards ES purposes. Single-shot detection sensitivity approaching the femtogram range and relative isotope abundance uncertainty better than 10% has been demonstrated using gadolinium isotopes as surrogate materials.

  15. Italy Highly Enriched Uranium and Plutonium Removals | National Nuclear

    National Nuclear Security Administration (NNSA)

    Security Administration Highly Enriched Uranium and Plutonium Removals March 24, 2014 Italy has been a global leader in nuclear nonproliferation, working with the United States since 1997 to eliminate more than 100 kilograms of highly enriched uranium (HEU) and separated plutonium. At the 2014 Nuclear Security Summit, the United States and Italy announced the successful removal of all eligible fresh HEU and plutonium from Italy. These shipments were completed via a joint effort between the

  16. NNSA Partnership Successfully Removes All Remaining HEU from Uzbekistan |

    National Nuclear Security Administration (NNSA)

    National Nuclear Security Administration Partnership Successfully Removes All Remaining HEU from Uzbekistan September 29, 2015 WASHINGTON, DC - Today, the Department of Energy's National Nuclear Security Administration (DOE/NNSA) announced the successful return of the final 5 kilograms (approximately 11 pounds) of highly enriched uranium (HEU) spent fuel from the IIN-3M "Foton" research reactor in Tashkent, Uzbekistan to Russia. This is the eighth shipment of HEU from Uzbekistan

  17. Secure Transportation of HEU in Romania

    SciTech Connect (OSTI)

    2009-07-06

    The National Nuclear Security Administration has announced the final shipments of Russian-origin highly enriched uranium (HEU) nuclear fuel from Romania. The material was removed and returned to Russia by air for storage at two secure nuclear facilities, making Romania the first country to remove all HEU since President Obama outlined his commitment to securing all vulnerable nuclear material around the world within four years. This was also the first time NNSA has shipped spent HEU by airplane, a development that will help accelerate efforts to meet the Presidents objective.

  18. EIS-0240: Disposition of Surplus Highly Enriched Uranium

    Broader source: Energy.gov [DOE]

    The Department proposes to eliminate the proliferation threat of surplus highly enriched uranium (HEU) by blending it down to low enriched uranium (LEU), which is not weapons-usable. The EIS assesses the disposition of a nominal 200 metric tons of surplus HEU. The Preferred Alternative is, where practical, to blend the material for use as LEU and use overtime, in commercial nuclear reactor field to recover its economic value. Material that cannot be economically recovered would be blended to LEU for disposal as low-level radioactive waste.

  19. HEU Minimization and the Reliable Supply of Medical Isotopes Nuclear

    National Nuclear Security Administration (NNSA)

    Security Summit: Fact Sheet | National Nuclear Security Administration HEU Minimization and the Reliable Supply of Medical Isotopes Nuclear Security Summit: Fact Sheet March 26, 2012 Molybdenum-99 (Mo-99) is used to produce technetium-99m (Tc-99m), a medical isotope that is used in about 100,000 diagnostic medical procedures globally every day. Today, Mo-99 is produced at aging facilities in Europe, Canada and South Africa primarily using highly-enriched uranium (HEU) - a weapons-usable

  20. Feasibility analyses for HEU to LEU fuel conversion of the LAUE Langivin Institute (ILL) High Flux Reactor (RHF).

    SciTech Connect (OSTI)

    Stevens, J.; Tentner. A.; Bergeron, A.; Nuclear Engineering Division

    2010-08-19

    The High Flux Reactor (RHF) of the Laue Langevin Institute (ILL) based in Grenoble, France is a research reactor designed primarily for neutron beam experiments for fundamental science. It delivers one of the most intense neutron fluxes worldwide, with an unperturbed thermal neutron flux of 1.5 x 10{sup 15} n/cm{sup 2}/s in its reflector. The reactor has been conceived to operate at a nuclear power of 57 MW but currently operates at 52 MW. The reactor currently uses a Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most worldwide research and test reactors have already started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on a mixture of uranium and molybdenum (UMo) is expected to allow the conversion of compact high performance reactors like the RHF. This report presents the results of reactor design, performance and steady state safety analyses for conversion of the RHF from the use of HEU fuel to the use of UMo LEU fuel. The objective of this work was to show that is feasible, under a set of manufacturing assumptions, to design a new RHF fuel element that could safely replace the HEU element currently used. The new proposed design has been developed to maximize performance, minimize changes and preserve strong safety margins. Neutronics and thermal-hydraulics models of the RHF have been developed and qualified by benchmark against experiments and/or against other codes and models. The models developed were then used to evaluate the RHF performance if LEU UMo were to replace the current HEU fuel 'meat' without any geometric change to the fuel plates. Results of these direct replacement analyses have shown a significant degradation of the RHF performance, in terms of both neutron flux and cycle length. Consequently, ANL and ILL have collaborated to investigate alternative designs. A promising candidate design has been selected and studied, increasing the total amount of fuel without changing the external plate dimensions by relocating the burnable poison. In this way, changes required in the fuel element are reasonably small. With this new design, neutronics analyses have shown that performance could be maintained at a high level: 2 day decrease of cycle length (to 47.5 days at 58.3 MW) and 1-2% decrease of brightness in the cold and hot sources in comparison to the current typical operation. In addition, studies have shown that the thermal-hydraulic and shutdown margins for the proposed LEU design would satisfy technical specifications.

  1. Why is weapons grade plutonium more hazardous to work with than highly enriched uranium?

    SciTech Connect (OSTI)

    Cournoyer, Michael E.; Costigan, Stephen A.; Schake, Bradley S.

    2015-08-01

    Highly Enriched Uranium and Weapons grade plutonium have assumed positions of dominant importance among the actinide elements because of their successful uses as explosive ingredients in nuclear weapons and the place they hold as key materials in the development of industrial use of nuclear power. While most chemists are familiar with the practical interest concerning HEU and WG Pu, fewer know the subtleties among their hazards. In this study, a primer is provided regarding the hazards associated with working with HEU and WG Pu metals and oxides. The care that must be taken to safely handle these materials is emphasized and the extent of the hazards is described. The controls needed to work with HEU and WG Pu metals and oxides are differentiated. Given the choice, one would rather work with HEU metal and oxides than WG Pu metal and oxides.

  2. Why is weapons grade plutonium more hazardous to work with than highly enriched uranium?

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Cournoyer, Michael E.; Costigan, Stephen A.; Schake, Bradley S.

    2015-08-01

    Highly Enriched Uranium and Weapons grade plutonium have assumed positions of dominant importance among the actinide elements because of their successful uses as explosive ingredients in nuclear weapons and the place they hold as key materials in the development of industrial use of nuclear power. While most chemists are familiar with the practical interest concerning HEU and WG Pu, fewer know the subtleties among their hazards. In this study, a primer is provided regarding the hazards associated with working with HEU and WG Pu metals and oxides. The care that must be taken to safely handle these materials is emphasizedmore » and the extent of the hazards is described. The controls needed to work with HEU and WG Pu metals and oxides are differentiated. Given the choice, one would rather work with HEU metal and oxides than WG Pu metal and oxides.« less

  3. HIGHLY ENRICHED URANIUM BLEND DOWN PROGRAM AT THE SAVANNAH RIVER SITE PRESENT AND FUTURE

    SciTech Connect (OSTI)

    Magoulas, V; Charles Goergen, C; Ronald Oprea, R

    2008-06-05

    The Department of Energy (DOE) and Tennessee Valley Authority (TVA) entered into an Interagency Agreement to transfer approximately 40 metric tons of highly enriched uranium (HEU) to TVA for conversion to fuel for the Browns Ferry Nuclear Power Plant. Savannah River Site (SRS) inventories included a significant amount of this material, which resulted from processing spent fuel and surplus materials. The HEU is blended with natural uranium (NU) to low enriched uranium (LEU) with a 4.95% 235U isotopic content and shipped as solution to the TVA vendor. The HEU Blend Down Project provided the upgrades needed to achieve the product throughput and purity required and provided loading facilities. The first blending to low enriched uranium (LEU) took place in March 2003 with the initial shipment to the TVA vendor in July 2003. The SRS Shipments have continued on a regular schedule without any major issues for the past 5 years and are due to complete in September 2008. The HEU Blend program is now looking to continue its success by dispositioning an additional approximately 21 MTU of HEU material as part of the SRS Enriched Uranium Disposition Project.

  4. U.S. transparency monitoring of HEU oxide conversion and blending to LEU hexafluoride at three Russian blending plants

    SciTech Connect (OSTI)

    Leich, D., LLNL

    1998-07-27

    The down-blending of Russian highly enriched uranium (HEU) takes place at three Russian gaseous centrifuge enrichment plants. The fluorination of HEU oxide and down-blending of HEU hexafluoride began in 1994, and shipments of low enriched uranium (LEU) hexafluoride product to the United States Enrichment Corporation (USEC) began in 1995 US transparency monitoring under the HEU Purchase Agreement began in 1996 and includes a permanent monitoring presence US transparency monitoring at these facilities is intended to provide confidence that HEU is received and down-blended to LEU for shipment to USEC The monitoring begins with observation of the receipt of HEU oxide shipments, including confirmation of enrichment using US nondestructive assay equipment The feeding of HEU oxide to the fluorination process and the withdrawal of HEU hexafluoride are monitored Monitoring is also conducted where the blending takes place and where shipping cylinders are filled with LEU product. A series of process and material accountancy documents are provided to US monitors.

  5. Research Reactor Preparations for the Air Shipment of Highly Enriched Uranium from Romania

    SciTech Connect (OSTI)

    K. J. Allen; I. Bolshinsky; L. L. Biro; M. E. Budu; N. V. Zamfir; M. Dragusin; C. Paunoiu; M. Ciocanescu

    2010-03-01

    In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation for conversion to low enriched uranium. The Institute for Nuclear Research at Pitesti (SCN Pitesti) shipped 30.1 kg of HEU fresh fuel pellets to Dimitrovgrad, Russia and the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH) shipped 23.7 kilograms of HEU spent fuel assemblies from the VVR S research reactor at Magurele, Romania, to Chelyabinsk, Russia. Both HEU shipments were coordinated by the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), were managed in Romania by the National Commission for Nuclear Activities Control (CNCAN), and were conducted in cooperation with the Russian Federation State Corporation Rosatom and the International Atomic Energy Agency. Both shipments were transported by truck to and from respective commercial airports in Romania and the Russian Federation and stored at secure nuclear facilities in Russia until the material is converted into low enriched uranium. These shipments resulted in Romania becoming the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the research reactor preparations and license approvals that were necessary to safely and securely complete these air shipments of nuclear fuel.

  6. AIR SHIPMENT OF HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL FROM ROMANIA AND LIBYA

    SciTech Connect (OSTI)

    Christopher Landers; Igor Bolshinsky; Ken Allen; Stanley Moses

    2010-07-01

    In June 2009 Romania successfully completed the worlds first air shipment of highly enriched uranium (HEU) spent nuclear fuel transported in Type B(U) casks under existing international laws and without special exceptions for the air transport licenses. Special 20-foot ISO shipping containers and cask tiedown supports were designed to transport Russian TUK 19 shipping casks for the Romanian air shipment and the equipment was certified for all modes of transport, including road, rail, water, and air. In December 2009 Libya successfully used this same equipment for a second air shipment of HEU spent nuclear fuel. Both spent fuel shipments were transported by truck from the originating nuclear facilities to nearby commercial airports, were flown by commercial cargo aircraft to a commercial airport in Yekaterinburg, Russia, and then transported by truck to their final destinations at the Production Association Mayak facility in Chelyabinsk, Russia. Both air shipments were performed under the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI). The Romania air shipment of 23.7 kg of HEU spent fuel from the VVR S research reactor was the last of three HEU fresh and spent fuel shipments under RRRFR that resulted in Romania becoming the 3rd RRRFR participating country to remove all HEU. Libya had previously completed two RRRFR shipments of HEU fresh fuel so the 5.2 kg of HEU spent fuel air shipped from the IRT 1 research reactor in December made Libya the 4th RRRFR participating country to remove all HEU. This paper describes the equipment, preparations, and license approvals required to safely and securely complete these two air shipments of spent nuclear fuel.

  7. The Role of COMSOL Toward a Low-Enriched Uranium Fuel Design for the High

    Office of Scientific and Technical Information (OSTI)

    Flux Isotope Reactor (Conference) | SciTech Connect Conference: The Role of COMSOL Toward a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor Citation Details In-Document Search Title: The Role of COMSOL Toward a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor Design and safety analyses are underway to convert the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) from a high-enriched uranium (HEU) fuel to a low-enriched uranium

  8. Conversion and Blending Facility highly enriched uranium to low enriched uranium as metal. Revision 1

    SciTech Connect (OSTI)

    1995-07-05

    The mission of this Conversion and Blending Facility (CBF) will be to blend surplus HEU metal and alloy with depleted uranium metal to produce an LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal.

  9. Microsoft PowerPoint - 7_GARY_LANGLIE_NMMSS 2014 NRC HEU Report Presentation 21April2014.pptx

    National Nuclear Security Administration (NNSA)

    Regulatory Commission Report to the U.S. Congress: High Enriched Uranium Gary Langlie, Nuclear Regulatory Commission Background  American Medical Isotopes Production Act of 2012 (AMIPA) - passed as part of the National Defense Authorization Act for FY13 (NDAA), which was signed into law on January 2, 2013  AMIPA's primary focus - to improve reliability of domestic medical isotope supplies without using high enriched uranium (HEU)* and to eventually eliminate HEU exports  Section 3175 of

  10. Parametric Evaluation of Active Neutron Interrogation for the Detection of Shielded Highly-Enriched Uranium in the Field

    SciTech Connect (OSTI)

    D. L. Chcihester; E. H. Seabury; S. J. Thompson; R. R. C. Clement

    2011-10-01

    Parametric studies using numerical simulations are being performed to assess the performance capabilities and limits of active neutron interrogation for detecting shielded highly enriched uranium (HEU). Varying the shield material, HEU mass, HEU depth inside the shield, and interrogating neutron source energy, the simulations account for both neutron and photon emission signatures from the HEU with resolution in both energy and time. The results are processed to represent different irradiation timing schemes and several different classes of radiation detectors, and evaluated using a statistical approach considering signal intensity over background. This paper describes the details of the modeling campaign and some preliminary results, weighing the strengths of alternative measurement approaches for the different irradiation scenarios.

  11. highly enriched uranium

    National Nuclear Security Administration (NNSA)

    and radioisotope supply capabilities of MURR and Nordion with General Atomics' selective gas extraction technology-which allows their low-enriched uranium (LEU) targets to remain...

  12. Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics

    SciTech Connect (OSTI)

    Stillman, J.; Feldman, E.; Foyto, L; Kutikkad, K; McKibben, J C; Peters, N.; Stevens, J.

    2012-09-01

    This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.

  13. A simple method for rapidly processing HEU from weapons returns

    SciTech Connect (OSTI)

    McLean, W. II; Miller, P.E.

    1994-01-01

    A method based on the use of a high temperature fluidized bed for rapidly oxidizing, homogenizing and down-blending Highly Enriched Uranium (HEU) from dismantled nuclear weapons is presented. This technology directly addresses many of the most important issues that inhibit progress in international commerce in HEU; viz., transaction verification, materials accountability, transportation and environmental safety. The equipment used to carry out the oxidation and blending is simple, inexpensive and highly portable. Mobile facilities to be used for point-of-sale blending and analysis of the product material are presented along with a phased implementation plan that addresses the conversion of HEU derived from domestic weapons and related waste streams as well as material from possible foreign sources such as South Africa or the former Soviet Union.

  14. High Accuracy U-235 Enrichment Verification Station for Low Enriched Uranium Alloys

    SciTech Connect (OSTI)

    Lillard, C. R.; Hayward, J. P.; Williamson, M. R.

    2012-06-07

    The Y-12 National Security Complex is playing a role in the U.S. High Performance Research Reactor (USHPRR) Conversion program sponsored by the U.S. National Nuclear Security Administration's Office of Global Threat Reduction. The USHPRR program has a goal of converting remaining U.S. reactors that continue to use highly enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. The USHPRR program is currently developing a LEU Uranium-Molybdenum (U-Mo) monolithic fuel for use in the U.S. high performance research reactors.Y-12 is supporting both the fuel development and fuel fabrication efforts by fabricating low enriched U-Mo foils from its own source material for irradiation experiments and for optimizing the fabrication process in support of scaling up the process to a commercial production scale. Once the new fuel is qualified, Y-12 will produce and ship U-Mo coupons with verified 19.75% +0.2% - 0.3% U-235 enrichment to be fabricated into fuel elements for the USHPRRs. Considering this small enrichment tolerance and the transition into HEU being set strictly at 20% U-235, a characterization system with a measurement uncertainty of less than or equal to 0.1% in enrichment is desired to support customer requirements and minimize production costs. Typical uncertainty for most available characterization systems today is approximately 1-5%; therefore, a specialized system must be developed which results in a reduced measurement uncertainty. A potential system using a High-Purity Germanium (HPGe) detector has been procured, and tests have been conducted to verify its capabilities with regards to the requirements. Using four U-Mo enrichment standards fabricated with complete isotopic and chemical characterization, infinite thickness and peak-ratio enrichment measurement methods have been considered for use. As a result of inhomogeneity within the U-Mo samples, FRAM, an isotopic analysis software, has been selected for initial testing. A systematic approach towards observing effects on FRAM's enrichment analysis has been conducted with regards to count and dead time.

  15. Highly Enriched Uranium Materials Facility

    National Nuclear Security Administration (NNSA)

    Appropriations Subcommittee, is shown some of the technology in the Highly Enriched Uranium Materials Facility by Warehousing and Transportation Operations Manager Byron...

  16. Estimation of the Performance of Multiple Active Neutron Interrogation Signatures for Detecting Shielded HEU

    SciTech Connect (OSTI)

    David L. Chichester; Scott J. Thompson; Scott M. Watson; James T. Johnson; Edward H. Seabury

    2012-10-01

    A comprehensive modeling study has been carried out to evaluate the utility of multiple active neutron interrogation signatures for detecting shielded highly enriched uranium (HEU). The modeling effort focused on varying HEU masses from 1 kg to 20 kg; varying types of shields including wood, steel, cement, polyethylene, and borated polyethylene; varying depths of the HEU in the shields, and varying engineered shields immediately surrounding the HEU including steel, tungsten, and cadmium. Neutron and gamma-ray signatures were the focus of the study and false negative detection probabilities versus measurement time were used as a performance metric. To facilitate comparisons among different approaches an automated method was developed to generate receiver operating characteristic (ROC) curves for different sets of model variables for multiple background count rate conditions. This paper summarizes results or the analysis, including laboratory benchmark comparisons between simulations and experiments. The important impact engineered shields can play towards degrading detectability and methods for mitigating this will be discussed.

  17. Highly Enriched Uranium Materials Facility | National Nuclear...

    National Nuclear Security Administration (NNSA)

    Gallery Jobs Apply for Our Jobs Our Jobs Working at NNSA Blog Home Highly Enriched Uranium Materials Facility Highly Enriched Uranium Materials Facility Congressmen tour Y-12...

  18. Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the USA and the Russian Federation has on the Domestic Uranium Mining, Conversion, and Enrichment Industries and the Ops of the Gaseous Diffusion

    Broader source: Energy.gov [DOE]

    The successful implementation of the HEU Agreement remains a high priority of the U.S. Government. The agreement also serves U.S. and Russian commercial interests. HEU Agreement deliveries are an...

  19. Overview of Russian HEU transparency issues

    SciTech Connect (OSTI)

    Kempf, C.R.; Bieniawski, A.

    1993-09-01

    The U.S. has signed an agreement with the Russian Federation for the purchase of 500 metric tons of highly-enriched uranium (HEU) taken from dismantled nuclear weapons. The HEU will be blended down to low-enriched uranium and will be transported to the U.S. to be used by fuel fabricators to make fuel for commercial nuclear power plants. Both the U.S. and Russia have been preparing to institute transparency measures to provide assurance that nonproliferation and arms control objectives specified in the agreement are met. This paper provides background information on the original agreement and on subsequent negotiations with the Russians, as well as discussion of technical aspects of developing transparency measures suited to the facilities and processes which are expected to be involved. Transparency has been defined as those agreed-upon measures which build confidence that arms control and non-proliferation objectives shared by the parties are met. Transparency is a departure from exhaustive, detailed arms control verification regimes of past agreements, which were based on a presumption of detecting transgressions as opposed to confirming compliance.

  20. HEU to LEU conversion and blending facility: Metal blending alternative to produce LEU oxide for disposal

    SciTech Connect (OSTI)

    1995-09-01

    US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. The nuclear material is converted to a form more proliferation- resistant than the original form. Blending HEU (highly enriched uranium) with less-enriched uranium to form LEU has been proposed as a disposition option. Five technologies are being assessed for blending HEU. This document provides data to be used in environmental impact analysis for the HEU-LEU disposition option that uses metal blending with an oxide waste product. It is divided into: mission and assumptions, conversion and blending facility descriptions, process descriptions and requirements, resource needs, employment needs, waste and emissions from plant, hazards discussion, and intersite transportation.

  1. Highly Enriched Uranium Materials Facility | Y-12 National Security...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Highly Enriched Uranium ... Highly Enriched Uranium Materials Facility HEUMF The Highly Enriched Uranium Materials Facility is our nation's central repository for highly enriched...

  2. Validation of ATR Fission Power Deposition Fraction in HEU and LEU Fuel Plates

    SciTech Connect (OSTI)

    G. S. Chang

    2008-09-01

    The Advanced Test Reactor (ATR) is a high power (250 MW), high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2s. Because of its high power and large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. A detailed plate-by-plate MCNP ATR full core model has been developed and validated for the low-enriched uranium (LEU) fuel conversion feasibility study. Using this model, an analysis has been performed to determine the LEU density and U-235 enrichment required in the fuel meat to yield equivalent K-eff versus effective full power days (EFPDs) between the HEU and LEU cores. This model has also been used to optimize U-235 content of the LEU core, minimizing the differences in K-eff and heat flux profile between the HEU and LEU cores at 115 MW total core power for 125 EFPDs. The LEU core conversion feasibility study evaluated foil type (U-10Mo) fuel with the LEU reference design of 19.7 wt% U-235 enrichment. The LEU reference design has a fixed fuel meat thickness of 0.330 mm and can sustain the same operating cycle length as the HEU fuel. Heat flux and fission power density are parameters that are proportional to the fraction of fission power deposited in fuel. Thus, the accurate determination of the fraction of fission power deposited in the fuel is important to ATR nuclear safety. In this work, a new approach was developed and validated, the Tally Fuel Cells Only (TFCO) method. This method calculates and compares the fission power deposition fraction between HEU and LEU fuel plates. Due to the high density of the U-10Mo LEU fuel, the fission ?-energy deposition fraction is 37.12%, which is larger than the HEUs ?-energy deposition fraction of 19.7%. As a result, the fuel decay heat cooling will need to be improved. During the power operation, the total fission energy (200 MeV per fission) deposition fraction of LEU and HEU are 90.9% and 89.1%, respectively.

  3. Establishing a Cost Basis for Converting the High Flux Isotope Reactor from High Enriched to Low Enriched Uranium Fuel

    SciTech Connect (OSTI)

    Primm, Trent; Guida, Tracey

    2010-02-01

    Under the auspices of the Global Threat Reduction Initiative Reduced Enrichment for Research and Test Reactors Program, the National Nuclear Security Administration /Department of Energy (NNSA/DOE) has, as a goal, to convert research reactors worldwide from weapons grade to non-weapons grade uranium. The High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL) is one of the candidates for conversion of fuel from high enriched uranium (HEU) to low enriched uranium (LEU). A well documented business model, including tasks, costs, and schedules was developed to plan the conversion of HFIR. Using Microsoft Project, a detailed outline of the conversion program was established and consists of LEU fuel design activities, a fresh fuel shipping cask, improvements to the HFIR reactor building, and spent fuel operations. Current-value costs total $76 million dollars, include over 100 subtasks, and will take over 10 years to complete. The model and schedule follows the path of the fuel from receipt from fuel fabricator to delivery to spent fuel storage and illustrates the duration, start, and completion dates of each subtask to be completed. Assumptions that form the basis of the cost estimate have significant impact on cost and schedule.

  4. DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

    SciTech Connect (OSTI)

    Cook, David Howard; Freels, James D; Ilas, Germina; Jolly, Brian C; Miller, James Henry; Primm, Trent; Renfro, David G; Sease, John D; Pinkston, Daniel

    2011-02-01

    This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

  5. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual report for FY 2009

    SciTech Connect (OSTI)

    Chandler, David; Freels, James D; Ilas, Germina; Miller, James Henry; Primm, Trent; Sease, John D; Guida, Tracey; Jolly, Brian C

    2010-02-01

    This report documents progress made during FY 2009 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Studies are reported of the application of a silicon coating to surrogates for spheres of uranium-molybdenum alloy. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. A description of the progress in developing a finite element thermal hydraulics model of the LEU core is provided.

  6. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

    SciTech Connect (OSTI)

    Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

    2009-03-01

    This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

  7. Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01

    An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

  8. 2005 Annual Health Physics Report for HEU Transparency Program

    SciTech Connect (OSTI)

    Radev, R

    2006-04-21

    During the 2005 calendar year, LLNL provided health physics support for the Highly Enriched Uranium Transparency Program (HEU-TP) in external and internal radiation protection and technical expertise into matters related to BDMS radioactive sources and Russian radiation safety regulatory compliance. For the calendar year 2005, there were 161 person-trips that required dose monitoring of the U.S. monitors. Of the 161 person-trips, 149 person-trips were SMVs and 12 person-trips were Transparency Monitoring Office (TMO) trips. Additionally, there were 11 monitoring visits by TMO monitors to facilities other than UEIE and 3 to UEIE itself. There were two monitoring visits (source changes) that were back to back with 16 monitors. Each of these concurring visits were treated as single person-trips for dosimetry purposes. Counted individually, there were 191 individual person-visits in 2005. The LLNL Safety Laboratories Division provided the dosimetry services for the HEU-TP monitors.

  9. ES-3100: A New Generation Shipping Container for Bulk Highly Enriched Uranium and Other Fissile Materials

    SciTech Connect (OSTI)

    Arbital, J.G.; Byington, G.A.; Tousley, D.R.

    2004-07-01

    The U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) is shipping bulk quantities of surplus fissile materials, primarily highly enriched uranium (HEU), over the next 15 to 20 years for disposition purposes. The U.S. Department of Transportation (DOT) specification 6M container is the package of choice for most of these shipments. However, the 6M does not conform to the Type B packaging requirements in the ''Code of Federal Regulations'' (10CFR71) and, for that reason, is being phased out for use in the secure transportation system of DOE. BWXT Y-12 is currently developing a package to replace the DOT 6M container for HEU disposition shipping campaigns. The new package is based on state-of-the-art, proven, and patented insulation technologies that have been successfully applied in the design of other packages. The new package, designated the ES-3100, will have a 50% greater capacity for HEU than the 6M and will be easier to use. Engineering analysis on the new package includes detailed dynamic impact finite element analysis (FEA). This analysis gives the ES-3100 a high probability of complying with regulatory requirements.

  10. Environmental Assessment for the Transportation of Highly Enriched Uranium from the Russian Federation for the Y-12 National Security Complex and Finding of No Significant Impact

    SciTech Connect (OSTI)

    2004-01-01

    The United States (U.S.) Department of Energy (DOE) proposes to transport highly enriched uranium (HEU) from Russia to a secure storage facility in Oak Ridge, TN. This proposed action would allow the United States and Russia to accelerate the disposition of excess nuclear weapons materials in the interest of promoting nuclear disarmament, strengthening nonproliferation, and combating terrorism. The HEU would be used for a non-weapons purpose in the U.S. as fuel in research reactors performing solely peaceful missions.

  11. Comparison and validation of HEU and LEU modeling results to HEU experimental benchmark data for the Massachusetts Institute of Technology MITR reactor.

    SciTech Connect (OSTI)

    Newton, T. H.; Wilson, E. H; Bergeron, A.; Horelik, N.; Stevens, J. (Nuclear Engineering Division); (MIT Nuclear Reactor Lab.)

    2011-03-02

    The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Towards this goal, comparisons of MCNP5 Monte Carlo neutronic modeling results for HEU and LEU cores have been performed. Validation of the model has been based upon comparison to HEU experimental benchmark data for the MITR-II. The objective of this work was to demonstrate a model which could represent the experimental HEU data, and therefore could provide a basis to demonstrate LEU core performance. This report presents an overview of MITR-II model geometry and material definitions which have been verified, and updated as required during the course of validation to represent the specifications of the MITR-II reactor. Results of calculations are presented for comparisons to historical HEU start-up data from 1975-1976, and to other experimental benchmark data available for the MITR-II Reactor through 2009. This report also presents results of steady state neutronic analysis of an all-fresh LEU fueled core. Where possible, HEU and LEU calculations were performed for conditions equivalent to HEU experiments, which serves as a starting point for safety analyses for conversion of MITR-II from the use of HEU fuel to the use of UMo LEU fuel.

  12. PREPARING THE HIGH FLUX ISOTOPE REACTOR FOR CONVERSION TO LOW ENRICHED URANIUM FUEL ? RETURN TO 100 MW

    SciTech Connect (OSTI)

    Smith, Kevin Arthur [ORNL; Primm, Trent [ORNL

    2009-01-01

    The feasibility of low-enriched uranium (LEU) fuel as a replacement for the current, high enriched uranium (HEU) fuel for the High Flux Isotope Reactor (HFIR) has been under study since 2006. Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting to LEU fuel requires returning the reactor power to 100 MW from 85 MW. The analyses required to up-rate the reactor power and the methods to perform these analyses are discussed. Comments regarding the regulatory approval process are provided along with a conceptual schedule.

  13. Analysis of HEU samples from the ULBA Metallurgical Plant

    SciTech Connect (OSTI)

    Gift, E.H.

    1995-05-01

    In early March 1994, eight highly enriched uranium (HEU) samples were collected from materials stored at the Ulba Metallurgical Plant in Oskamen (Ust Kamenogorsk), Kazakhstan. While at the plant site, portions of four samples were dissolved and analyzed by mass spectrograph at the Ulba analytical laboratory by Ulba analysts. Three of these mass spectrograph solutions and the eight HEU samples were subsequently delivered to the Y-12 Plant for complete chemical and isotopic analyses. Chemical forms of the eight samples were uranium metal chips, U0{sub 2} powder, uranium/beryllium oxide powder, and uranium/beryllium alloy rods. All were declared by the Ulba plant to have a uranium assay of {approximately}90 wt % {sup 235}U. The uranium/beryllium powder and alloy samples were also declared to range from about 8 to 28 wt % uranium. The chemical and uranium isotopic analyses done at the Y-12 Plant confirm the Ulba plant declarations. All samples appear to have been enriched using some reprocessed uranium, probably from recovery of uranium from plutonium production reactors. As a result, all samples contain some {sup 236}U and {sup 232}U and have small but measurable quantities of plutonium. This plutonium could be the result of either contamination carried over from the enrichment process or cross-contamination from weapons material. It is not the result of direct reactor exposure. Neither the {sup 232}U nor the plutonium concentrations are sufficiently high to provide a significant industrial health hazard. Both are well within established or proposed acceptance criteria for storage at Y-12. The trace metal analyses showed that, with the exception of beryllium, there are no trace metals in any of these HEU samples that pose a significant health hazard.

  14. Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

    SciTech Connect (OSTI)

    Renfro, David G; Cook, David Howard; Freels, James D; Griffin, Frederick P; Ilas, Germina; Sease, John D; Chandler, David

    2012-03-01

    This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

  15. Design Study for a Low-enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2007

    SciTech Connect (OSTI)

    Primm, Trent; Ellis, Ronald James; Gehin, Jess C; Ilas, Germina; Miller, James Henry; Sease, John D

    2007-11-01

    This report documents progress made during fiscal year 2007 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low enriched uranium fuel (LEU). Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. A high volume fraction U/Mo-in-Al fuel could attain the same neutron flux performance as with the current, HEU fuel but materials considerations appear to preclude production and irradiation of such a fuel. A diffusion barrier would be required if Al is to be retained as the interstitial medium and the additional volume required for this barrier would degrade performance. Attaining the high volume fraction (55 wt. %) of U/Mo assumed in the computational study while maintaining the current fuel plate acceptance level at the fuel manufacturer is unlikely, i.e. no increase in the percentage of plates rejected for non-compliance with the fuel specification. Substitution of a zirconium alloy for Al would significantly increase the weight of the fuel element, the cost of the fuel element, and introduce an as-yet untried manufacturing process. A monolithic U-10Mo foil is the choice of LEU fuel for HFIR. Preliminary calculations indicate that with a modest increase in reactor power, the flux performance of the reactor can be maintained at the current level. A linearly-graded, radial fuel thickness profile is preferred to the arched profile currently used in HEU fuel because the LEU fuel media is a metal alloy foil rather than a powder. Developments in analysis capability and nuclear data processing techniques are underway with the goal of verifying the preliminary calculations of LEU flux performance. A conceptual study of the operational cost of an LEU fuel fabrication facility yielded the conclusion that the annual fuel cost to the HFIR would increase significantly from the current, HEU fuel cycle. Though manufacturing can be accomplished with existing technology, several engineering proof-of-principle tests would be required. The RERTR program is currently conducting a series of generic fuel qualification tests at the Advanced Test Reactor. A review of these tests and a review of the safety basis for the current, HEU fuel cycle led to the identification of a set of HFIR-specific fuel qualification tests. Much additional study is required to formulate a HFIR-specific fuel qualification plan from this set. However, one such test - creating a graded fuel profile across a flat foil - has been initiated with promising results.

  16. Highly Enriched Uranium Disposition | National Nuclear Security...

    National Nuclear Security Administration (NNSA)

    the economic value of the material by using the resulting LEU as nuclear reactor fuel. ... HEU from Russian nuclear weapons into LEU used as fuel in U.S. commercial power reactors. ...

  17. Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania

    SciTech Connect (OSTI)

    K. J. Allen; I. Bolshinsky; L. L. Biro; M. E. Budu; N. V. Zamfir; M. Dragusin

    2010-07-01

    Romania safely air shipped 23.7 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel from the VVR S research reactor at Magurele, Romania, to the Russian Federation in June 2009. This was the worlds first air shipment of spent nuclear fuel transported in a Type B(U) cask under existing international laws without special exceptions for the air transport licenses. This shipment was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in cooperation with the Romania National Commission for Nuclear Activities Control (CNCAN), the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH), and the Russian Federation State Corporation Rosatom. The shipment was transported by truck to and from the respective commercial airports in Romania and the Russian Federation and stored at a secure nuclear facility in Russia where it will be converted into low enriched uranium. With this shipment, Romania became the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the work, equipment, and approvals that were required to complete this spent fuel air shipment.

  18. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2006

    SciTech Connect (OSTI)

    Primm, R. T.; Ellis, R. J.; Gehin, J. C.; Clarno, K. T.; Williams, K. A.; Moses, D. L.

    2006-11-01

    Neutronics and thermal-hydraulics studies show that, for equivalent operating power [85 MW(t)], a low-enriched uranium (LEU) fuel cycle based on uranium-10 wt % molybdenum (U-10Mo) metal foil with radially, continuously graded fuel meat thickness results in a 15% reduction in peak thermal flux in the beryllium reflector of the High Flux Isotope Reactor (HFIR) as compared to the current highly enriched uranium (HEU) cycle. The uranium-235 content of the LEU core is almost twice the amount of the HEU core when the length of the fuel cycle is kept the same for both fuels. Because the uranium-238 content of an LEU core is a factor of 4 greater than the uranium-235 content, the LEU HFIR core would weigh 30% more than the HEU core. A minimum U-10Mo foil thickness of 84 ?m is required to compensate for power peaking in the LEU core although this value could be increased significantly without much penalty. The maximum U-10Mo foil thickness is 457?m. Annual plutonium production from fueling the HFIR with LEU is predicted to be 2 kg. For dispersion fuels, the operating power for HFIR would be reduced considerably below 85 MW due to thermal considerations and due to the requirement of a 26-d fuel cycle. If an acceptable fuel can be developed, it is estimated that $140 M would be required to implement the conversion of the HFIR site at Oak Ridge National Laboratory from an HEU fuel cycle to an LEU fuel cycle. To complete the conversion by fiscal year 2014 would require that all fuel development and qualification be completed by the end of fiscal year 2009. Technological development areas that could increase the operating power of HFIR are identified as areas for study in the future.

  19. Overview of transparency issues under the US-Russian highly enriched uranium purchase agreement

    SciTech Connect (OSTI)

    Bieniawski, A.J.; Dougherty, D.R.

    1995-12-31

    The US has signed an Agreement with the Russian Federation for the purchase of 500 metric tons of highly enriched uranium (HEU) derived from dismantled Russian nuclear weapons. The BEU will be blended down to low-enriched uranium (LEU) in Russia and will be transported to the US to be used by fuel Fabricators to make fuel for commercial nuclear power plants. Both the United States and Russia have been preparing to institute transparency measures to provide confidence that the nonproliferation, physical protection, and material control and accounting requirements specified in the Agreement are met. This paper provides a background on the Agreement and subsequent on-going negotiations to develop transparency measures suited to the facilities and processes which are expected to be involved.

  20. Ion-induced gammas for photofission interrogation of HEU.

    SciTech Connect (OSTI)

    Doyle, Barney Lee (Sandia National Laboratories, Albuquerque, NM); Antolak, Arlyn J.; Morse, Daniel H.; Provencio, Paula Polyak (Sandia National Laboratories, Albuquerque, NM)

    2006-03-01

    High-energy photons and neutrons can be used to actively interrogate for heavily shielded special nuclear material (SNM), such as HEU (highly enriched uranium), by detecting prompt and/or delayed induced fission signatures. In this work, we explore the underlying physics for a new type of photon source that generates high fluxes of mono-energetic gamma-rays from low-energy (<500 keV) proton-induced nuclear reactions. The characteristic energies (4- to 18-MeV) of the gamma-rays coincide with the peak of the photonuclear cross section. The source could be designed to produce gamma-rays of certain selected energies, thereby improving the probability of detecting shielded HEU or providing a capability to determine enrichment inside sealed containers. The fundamental physics of such an interrogation source were studied in this LDRD through scaled ion accelerator experiments and radiation transport modeling. The data were used to assess gamma and neutron yields, background, and photofission-induced signal levels from several (p,{gamma}) target materials under consideration.

  1. TREAT Transient Analysis Benchmarking for the HEU Core

    SciTech Connect (OSTI)

    Kontogeorgakos, D. C.; Connaway, H. M.; Wright, A. E.

    2014-05-01

    This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to benchmark the transient calculations against temperature-limited transients performed in the final operating HEU TREAT core configuration. The MCNP code was used to evaluate steady-state neutronics behavior, and the point kinetics code TREKIN was used to determine core power and energy during transients. The first part of the benchmarking process was to calculate with MCNP all the neutronic parameters required by TREKIN to simulate the transients: the transient rod-bank worth, the prompt neutron generation lifetime, the temperature reactivity feedback as a function of total core energy, and the core-average temperature and peak temperature as a functions of total core energy. The results of these calculations were compared against measurements or against reported values as documented in the available TREAT reports. The heating of the fuel was simulated as an adiabatic process. The reported values were extracted from ANL reports, intra-laboratory memos and experiment logsheets and in some cases it was not clear if the values were based on measurements, on calculations or a combination of both. Therefore, it was decided to use the term reported values when referring to such data. The methods and results from the HEU core transient analyses will be used for the potential LEU core configurations to predict the converted (LEU) cores performance.

  2. EA-1977: Acceptance and Disposition of Used Nuclear Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany

    Broader source: Energy.gov [DOE]

    This environmental assessment (EA) will evaluate the potential environmental impacts of a DOE proposal to accept used nuclear fuel from the Federal Republic of Germany at DOEs Savannah River Site (SRS) for processing and disposition. This used nuclear fuel is composed of kernels containing thorium and U.S.-origin highly enriched uranium (HEU) embedded in small graphite spheres that were irradiated in nuclear reactors used for research and development purposes.

  3. Subcritical Measurements Multiple HEU Metal Castings

    SciTech Connect (OSTI)

    Mihalczo, John T [ORNL] [ORNL; Archer, Daniel E [ORNL] [ORNL; Wright, Michael C [ORNL] [ORNL

    2008-01-01

    Experiments with the standard annular highly enriched uranium (HEU) metal castings at Y-12 were performed in which up to 5 castings ({approx}90kg) were assembled in a tightly packed array with minimal spacing between castings. The fission chain multiplication process was initiated by a time tagged {sup 253}Cf spontaneously fissioning neutron source or time and directionally tagged neutrons from a small portable DT neutron generator and the prompt neutron time behavior measured with plastic scintillation detectors sensitive the fast neutron (>1 MeV) and gamma ray without distinction. These experiments were performed to provide data to benchmark methods for the calculation of the prompt neutron time behavior. Previous measurements with a single casting have been reported. This paper presents the experimental results for multiple castings. The prompt time decay was obtained by time coincidence correlation measurements between the detectors and the time tagged neutron source emission (equivalent to randomly pulsed neutron measurements) and between pairs of plastic scintillation detectors (equivalent to a 2-detector Rossi-alpha measurement). These standard HEU storage castings at the Y-12 plant had 5.000-in-OD, 3.500-in-ID, masses between 17,636 and 17,996 g, impurity content of 992 ppm, density of 18.75 g/cm{sup 3} and average enrichment of 93.16 wt % {sup 235}U. The castings were in tight fitting 025-in.-thick, 8.0-in-high stainless steel (SS-304) cylindrical cans for contamination control which were 8.0 in high. One can had an inside diameter of 3.0 in so that the Cf source could be located on the axes of this casting. Four 1 x 1 x 6 in plastic scintillators with the long dimension perpendicular to axes of the castings and adjacent to the outer surface of the casting cans were used. The detectors were enclosed in 1/4.-in.-thick lead shields on four 1 x 6 surfaces and on the 1 x 1 surface. The small surface of the lead shield was adjacent to the steel table. The photograph of the assembly of 5 casting with the detectors is given in Figure 1. The DT generator can be seen in the back with the time tagged neutrons aligned symmetrically around a line drawn from the generator through target spot and the center of the casting in the annular can seen in the foreground.

  4. Planning the HEU to LEU Transition for the NBSR

    SciTech Connect (OSTI)

    Hanson, A.L.; Diamond, D.

    2011-09-12

    A study has been carried out to understand how the NIST research reactor (NBSR) might be converted from using high-enriched uranium (HEU) to using low-enriched uranium (LEU) fuel. An LEU fuel design had previously been determined which provides an equilibrium core with the desirable fuel cycle length - a very important parameter for maintaining the experimental, scientific program supported by the NBSR. In the present study two options for getting to the equilibrium state are considered. One option starts with the loading of an entire core of fresh fuel. This was determined to be unacceptable. The other option makes use of the current fuel management scheme wherein four fresh fuel elements are loaded at the beginning of each cycle. However, it is shown that without some alterations to the fuel cycle, none of the transition cores containing both HEU and LEU fuel have sufficient excess reactivity to operate the reactor for the optimum length. It was determined that operating the first mixed cycle for a sufficiently reduced length of time provides the excess reactivity which enables subsequent cycles to be run for the desired number of days.

  5. Planning the HEU to LEU Transition for the NBSR

    SciTech Connect (OSTI)

    Hanson, A.L.; Diamond, D.

    2011-10-24

    A study has been carried out to understand how the NIST research reactor (NBSR) might be converted from using high-enriched uranium (HEU) to using low-enriched uranium (LEU) fuel. An LEU fuel design had previously been determined which provides an equilibrium core with the desirable fuel cycle lengtha very important parameter for maintaining the experimental, scientific program supported by the NBSR. In the present study two options for getting to the equilibrium state are considered. One option starts with the loading of an entire core of fresh fuel. This was determined to be unacceptable. The other option makes use of the current fuel management scheme wherein four fresh fuel elements are loaded at the beginning of each cycle. However, it is shown that without some alterations to the fuel cycle, none of the transition cores containing both HEU and LEU fuel have sufficient excess reactivity to enable reactor operation for the required amount of time. It was determined that operating the first mixed cycle for a sufficiently reduced length of time provides the excess reactivity which enables subsequent transition cycles to be run for the desired number of days.

  6. Prompt Neutron Time Decay in Single HEU and DU Metal Annular Storage Castings

    SciTech Connect (OSTI)

    Pena, Kirsten E [ORNL] [ORNL; McConchie, Seth M [ORNL] [ORNL; Mihalczo, John T [ORNL] [ORNL

    2010-01-01

    Previous measurements of highly enriched uranium (HEU) storage castings performed by Oak Ridge National Laboratory (ORNL) at the Y-12 National Security Complex showed a prompt neutron time decay that is not exponential. These measurements showed that multiple time constants originating from multiplication, time-of-flight, scattering in the assembly and room return could be associated with this prompt neutron decay. In this work, the contribution not associated with neutron multiplication was investigated via measurements with a depleted uranium (DU) casting. The measurements at ORNL used an annular (5.0-in OD, 3.5-in ID, 6.0-in H) DU casting with a time-tagged 252Cf source, centered vertically on the axis, and four closely coupled 1 1 6-in.-long plastic scintillators with -in.- thick lead shielding adjacent to the outer surface of the casting. This setup was identical to the configuration used in the previously performed measurements with HEU castings at Y-12. The time correlation between fission events and detections in the plastic scintillators was measured, as well as the time distribution of coincidences between multiple detectors within a 512-ns time window. The measurement results were then compared to MCNP-PoliMi calculations and the previous HEU measurements. Time constants from decay fits to the HEU and DU data were compared to characterize the contributions resulting from multiplication, time-of-flight, and scattering.

  7. Highly Enriched Uranyl Nitrate in Annular Tanks with Concrete Reflection: 1 x 3 Line Array of Nested Pairs of Tanks

    SciTech Connect (OSTI)

    James Cleaver; John D. Bess; Nathan Devine; Fitz Trumble

    2009-09-01

    A series of seven experiments were performed at the Rocky Flats Critical Mass Laboratory beginning in August, 1980 (References 1 and 2). Highly enriched uranyl nitrate solution was introduced into a 1-3 linear array of nested stainless steel annular tanks. The tanks were inside a concrete enclosure, with various moderator and absorber materials placed inside and/or between the tanks. These moderators and absorbers included boron-free concrete, borated concrete, borated plaster, and cadmium. Two configurations included placing bottles of highly enriched uranyl nitrate between tanks externally. Another experiment involved nested hemispheres of highly enriched uranium placed between tanks externally. These three configurations are not evaluated in this report. The experiments evaluated here are part of a series of experiments, one set of which is evaluated in HEU-SOL-THERM-033. The experiments in this and HEU-SOL-THERM-033 were performed similarly. They took place in the same room and used the same tanks, some of the same moderators and absorbers, some of the same reflector panels, and uranyl nitrate solution from the same location. There are probably additional similarities that existed that are not identified here. Thus, many of the descriptions in this report are either the same or similar to those in the HEU-SOL-THERM-033 report. Seventeen configurations (sixteen of which were critical) were performed during seven experiments; six of those experiments are evaluated here with thirteen configurations. Two configurations were identical, except for solution height, and were conducted to test repeatability. The solution heights were averaged and the two were evaluated as one configuration, which gives a total of twelve evaluated configurations. One of the seventeen configurations was subcritical. Of the twelve critical configurations evaluated, nine were judged as acceptable as benchmarks.

  8. Quadrilateral Cooperation on High-density Low-enriched Uranium...

    National Nuclear Security Administration (NNSA)

    Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel Production: Fact Sheet ... Fact Sheets Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel ... ...

  9. Highly Enriched Uranium Materials Facility, Major Design Changes...

    Energy Savers [EERE]

    Highly Enriched Uranium Materials Facility, Major Design Changes Late...Lessons Learned Report, NNSA, Dec 2010 Highly Enriched Uranium Materials Facility, Major Design Changes...

  10. Coordination Between the HEU Transparency Program and the Material Protection, Control and Accountability Program

    SciTech Connect (OSTI)

    Glaser, J.; Hernandez, J.; Dougherty, D.; Bieniawski, A.; Cahalane, P.; Mastal, E.

    2000-06-30

    DOE sponsored programs such as Material Protection Control and Accountability (MPC&A) and implementation of the Highly-Enriched Uranium (HEU) Transparency Program send US personnel into Russian nuclear facilities and receive Russian representatives from these programs. While there is overlap in the Russian nuclear facilities visited by these two programs, there had not been any formal mechanism to share information between them. Recently, an MPC&A/HEU Working Group was developed to facilitate the sharing of appropriate information and to address concerns expressed by Minatom and Russian facility personnel such as US visit scheduling conflicts. This paper discusses the goals of the Working Group and ways it has helped to allow the programs to work more efficiently with the Russian facilities.

  11. Fissile Mass Flow Monitor Implementation for Transparency in HEU Blenddown at the URAL Electrochemical Integrated Plant (UEIP) in Novouralsk

    SciTech Connect (OSTI)

    March-Leuba, J.; Mastal, E.; Powell, D.; Sumner, J.; Uckan, T.; Vines, B.

    1999-07-25

    The Oak Ridge National Laboratory (ORNL) Fissile Mass Flow Monitor (FMFM) was deployed at the Ural Electrochemical Integrated Plant (UEIP) highly enriched uranium (HEU) blending facility in January and February 1999 at Novouralsk in Russia for the DOE HEU Transparency Program. The FMFM provides unattended monitoring of the fissile mass flow of the uranium hexafluoride (UF{sub 6}) gas in the process lines of HEU, the low enriched uranium (LEU) blend stock, and the product LEU (P-LEU) of the blending tee non-intrusively. To do this, uranium-235 (U-235) fissions are induced in the UF{sub 6} by a thermalized and modulated californium-252 (Cf-252) neutron source placed on each process line. A set of detectors, located downstream of source, measure delayed gamma rays emitted by the resulting fission fragments. The observed delay in the time correlated measurement between the source and the detector signal provides the velocity of UF{sub 6} and its amplitude is related to the U- 235 content in UF{sub 6}. An on-line computer controls the source modulator, processes the collected detector data, and displays the results. The UEIP Main and the Reserved process lines were implemented with minor modifications. The FMFM monitors the HEU blending operation by measuring UF{sub 6} flows in the process blending lines, and the traceability of the HEU flow from the blend point to the P-LEU. The detail operational characteristics of the FMFM software (FM2) and the measurement methodology used are presented.

  12. highly enriched uranium | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    highly enriched uranium NNSA deputy administrator travels to Ukraine Earlier this month, Deputy Administrator for Defense Nuclear Nonproliferation Anne Harrington traveled to Ukraine to celebrate the 20th anniversary of the Science and Technology Center in Ukraine (STCU) and visit the Neutron Source Facility at the Kharkiv Institute of Physics and Technology. The U... DOE/NNSA Successfully Establishes Uranium Lease and Takeback Program to Support Critical Medical Isotope Production In January

  13. Mexico HEU Removal: Fact Sheet | National Nuclear Security Administrat...

    National Nuclear Security Administration (NNSA)

    Converting reactors from the use of HEU to LEU fuel and removing the remaining HEU prevents nuclear proliferation by permanently reducing the amount of HEU used in civilian ...

  14. RELAP5 model of the high flux isotope reactor with low enriched fuel thermal flux profiles

    SciTech Connect (OSTI)

    Banfield, J.; Mervin, B.; Hart, S.; Ritchie, J.; Walker, S.; Ruggles, A.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee Knoxville, Knoxville, TN 37996-2300 (United States)

    2012-07-01

    The High Flux Isotope Reactor (HFIR) currently uses highly enriched uranium (HEU) fabricated into involute-shaped fuel plates. It is desired that HFIR be able to use low enriched uranium (LEU) fuel while preserving the current performance capability for its diverse missions in material irradiation studies, isotope production, and the use of neutron beam lines for basic research. Preliminary neutronics and depletion simulations of HFIR with LEU fuel have arrived to feasible fuel loadings that maintain the neutronics performance of the reactor. This article illustrates preliminary models developed for the analysis of the thermal-hydraulic characteristics of the LEU core to ensure safe operation of the reactor. The beginning of life (BOL) LEU thermal flux profile has been modeled in RELAP5 to facilitate steady state simulation of the core cooling, and of anticipated and unanticipated transients. Steady state results are presented to validate the new thermal power profile inputs. A power ramp, slow depressurization at the outlet, and flow coast down transients are also evaluated. (authors)

  15. Reactor Physics Measurements and Benchmark Specifications for Oak Ridge Highly Enriched Uranium Sphere (ORSphere)

    SciTech Connect (OSTI)

    Marshall, Margaret A.

    2014-11-04

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an effort to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. Additionally, various material reactivity worths, the surface material worth coefficient, the delayed neutron fraction, the prompt neutron decay constant, relative fission density, and relative neutron importance were all measured. The critical assembly, material reactivity worths, the surface material worth coefficient, and the delayed neutron fraction were all evaluated as benchmark experiment measurements. The reactor physics measurements are the focus of this paper; although for clarity the critical assembly benchmark specifications are briefly discussed.

  16. Reactor Physics Measurements and Benchmark Specifications for Oak Ridge Highly Enriched Uranium Sphere (ORSphere)

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Marshall, Margaret A.

    2014-11-04

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an effort to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with themore » GODIVA I experiments. Additionally, various material reactivity worths, the surface material worth coefficient, the delayed neutron fraction, the prompt neutron decay constant, relative fission density, and relative neutron importance were all measured. The critical assembly, material reactivity worths, the surface material worth coefficient, and the delayed neutron fraction were all evaluated as benchmark experiment measurements. The reactor physics measurements are the focus of this paper; although for clarity the critical assembly benchmark specifications are briefly discussed.« less

  17. U. S. Department of Energy Savannah River Operations Office - German HEU

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Project Documents DOE Logo Search Home / German HEU Project Documents German HEU Project Live Broadcasts None currently scheduled. Past Meetings February 4th 2016 Department of Energy Meeting: VIDEO February 4th 2016 Department of Energy Meeting: Meeting Minutes February 4th 2016 Department of Energy Meeting: Presentation Slides June 24, 2014 German HEU Environmental Assessment Meeting Transcript June 24, 2014 Potential Acceptance & Disposition of Gernam Pebble Bed Reactor Highly

  18. Mexico HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Apply for Our Jobs Our Jobs Working at NNSA Blog Home content Four-Year Plan Mexico HEU Removal Mexico HEU Removal Location Mexico United States 24 24' 35.298" N, 102...

  19. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    SciTech Connect (OSTI)

    Daily, Charles R.

    2015-10-01

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  20. Gamma/neutron time-correlation for special nuclear material detection Active stimulation of highly enriched uranium

    SciTech Connect (OSTI)

    Paff, Marc G.; Monterial, Mateusz; Marleau, Peter; Kiff, Scott; Nowack, Aaron; Clarke, Shaun D.; Pozzi, Sara A.

    2014-06-21

    A series of simulations and experiments were undertaken to explore and evaluate the potential for a novel new technique for fissile material detection and characterization, the timecorrelated pulse-height (TCPH) method, to be used concurrent with active stimulation of potential nuclear materials. In previous work TCPH has been established as a highly sensitive method for the detection and characterization of configurations of fissile material containing Plutonium in passive measurements. By actively stimulating fission with the introduction of an external radiation source, we have shown that TCPH is also an effective method of detecting and characterizing configurations of fissile material containing Highly Enriched Uranium (HEU). The TCPH method is shown to be robust in the presence of the proper choice of external radiation source. An evaluation of potential interrogation sources is presented.

  1. Neutronics Analyses of the Minimum Original HEU TREAT Core

    SciTech Connect (OSTI)

    Kontogeorgakos, D.; Connaway, H.; Yesilyurt, G.; Wright, A.

    2014-04-01

    This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to validate the MCNP model of the TREAT reactor with the well-documented measurements which were taken during the start-up and early operation of TREAT. Furthermore, the effect of carbon graphitization was also addressed. The graphitization level was assumed to be 100% (ANL/GTRI/TM-13/4). For this purpose, a set of experiments was chosen to validate the TREAT MCNP model, involving the approach to criticality procedure, in-core neutron flux measurements with foils, and isothermal temperature coefficient and temperature distribution measurements. The results of this study extended the knowledge base for the TREAT MCNP calculations and established the credibility of the MCNP model to be used in the core conversion feasibility analysis.

  2. Gamma/neutron time-correlation for special nuclear material detection – Active stimulation of highly enriched uranium

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Paff, Marc G.; Monterial, Mateusz; Marleau, Peter; Kiff, Scott; Nowack, Aaron; Clarke, Shaun D.; Pozzi, Sara A.

    2014-06-21

    A series of simulations and experiments were undertaken to explore and evaluate the potential for a novel new technique for fissile material detection and characterization, the timecorrelated pulse-height (TCPH) method, to be used concurrent with active stimulation of potential nuclear materials. In previous work TCPH has been established as a highly sensitive method for the detection and characterization of configurations of fissile material containing Plutonium in passive measurements. By actively stimulating fission with the introduction of an external radiation source, we have shown that TCPH is also an effective method of detecting and characterizing configurations of fissile material containing Highlymore » Enriched Uranium (HEU). The TCPH method is shown to be robust in the presence of the proper choice of external radiation source. An evaluation of potential interrogation sources is presented.« less

  3. REMOVAL OF SOLIDS FROM HIGHLY ENRICHED URANIUM SOLUTIONS USING THE H-CANYON CENTRIFUGE

    SciTech Connect (OSTI)

    Rudisill, T; Fernando Fondeur, F

    2009-01-15

    Prior to the dissolution of Pu-containing materials in HB-Line, highly enriched uranium (HEU) solutions stored in Tanks 11.1 and 12.2 of H-Canyon must be transferred to provide storage space. The proposed plan is to centrifuge the solutions to remove solids which may present downstream criticality concerns or cause operational problems with the 1st Cycle solvent extraction due to the formation of stable emulsions. An evaluation of the efficiency of the H-Canyon centrifuge concluded that a sufficient amount (> 90%) of the solids in the Tank 11.1 and 12.2 solutions will be removed to prevent any problems. We based this conclusion on the particle size distribution of the solids isolated from samples of the solutions and the calculation of particle settling times in the centrifuge. The particle size distributions were calculated from images generated by scanning electron microscopy (SEM). The mean particle diameters for the distributions were 1-3 {micro}m. A significant fraction (30-50%) of the particles had diameters which were < 1 {micro}m; however, the mass of these solids is insignificant (< 1% of the total solids mass) when compared to particles with larger diameters. It is also probable that the number of submicron particles was overestimated by the software used to generate the particle distribution due to the morphology of the filter paper used to isolate the solids. The settling times calculated for the H-Canyon centrifuge showed that particles with diameters less than 1 to 0.5 {micro}m will not have sufficient time to settle. For this reason, we recommend the use of a gelatin strike to coagulate the submicron particles and facilitate their removal from the solution; although we have no experimental basis to estimate the level of improvement. Incomplete removal of particles with diameters < 1 {micro}m should not cause problems during purification of the HEU in the 1st Cycle solvent extraction. Particles with diameters > 1 {micro}m account for > 99% of the solid mass and will be efficiently removed by the centrifuge; therefore, the formation of emulsions during solvent extraction operations is not an issue. Under the current processing plan, the solutions from Tanks 11.1 and 12.2 will be transferred to the enriched uranium storage (EUS) tank following centrifugation. The solution from Tanks 11.1 and 12.2 may remain in the EUS tank for an extended time prior to purification. The effects of extended storage on the solution were not evaluated as part of this study.

  4. Neutronic Analyses for HEU to LEU fuel conversion of the Massachusetts Institute of Technology.

    SciTech Connect (OSTI)

    Wilson, E. H.; Newton, T. H.; Bergeron, A.; Horelik, N.; Stevens, J. G

    2011-03-02

    The Massachusetts Institute of Technology (MIT) reactor (MITR-II), based in Cambridge, Massachusetts, is a research reactor designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on a mixture of uranium and molybdenum (UMo) is expected to allow the conversion of compact high performance reactors like the MITR-II. This report presents the results of steady state neutronic safety analyses for conversion of MITR-II from the use of HEU fuel to the use of U-Mo LEU fuel. The objective of this work was to demonstrate that the safety analyses meet current requirements for an LEU core replacement of MITR-II.

  5. Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.

    SciTech Connect (OSTI)

    Wilson, E.H.; Horelik, N.E.; Dunn, F.E.; Newton, T.H., Jr.; Hu, L.; Stevens, J.G.

    2012-04-04

    The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Toward this goal, core geometry and power distributions are presented. Distributions of power are calculated for LEU cores depleted with MCODE using an MCNP5 Monte Carlo model. The MCNP5 HEU and LEU MITR models were previously compared to experimental benchmark data for the MITR-II. This same model was used with a finer spatial depletion in order to generate power distributions for the LEU cores. The objective of this work is to generate and characterize a series of fresh and depleted core peak power distributions, and provide a thermal hydraulic evaluation of the geometry which should be considered for subsequent thermal hydraulic safety analyses.

  6. ORSPHERE: PHYSICS MEASUREMENTS FOR BARE, HEU(93.2)-METAL SPHERE

    SciTech Connect (OSTI)

    Margaret A. Marshall

    2014-03-01

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an attempt to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s (HEU-MET-FAST-001). The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. “The very accurate description of this sphere, as assembled, establishes it as an ideal benchmark for calculational methods and cross-section data files” (Reference 1). While performing the ORSphere experiments care was taken to accurately document component dimensions (±0.0001 inches), masses (±0.01 g), and material data. The experiment was also set up to minimize the amount of structural material in the sphere proximity. Two, correlated spheres were evaluated and judged to be acceptable as criticality benchmark experiments. This evaluation is given in HEU-MET-FAST-100. The second, smaller sphere was used for additional reactor physics measurements. Worth measurements (Reference 1, 2, 3 and 4), the delayed neutron fraction (Reference 3, 4 and 5) and surface material worth coefficient (Reference 1 and 2) are all measured and judged to be acceptable as benchmark data. The prompt neutron decay (Reference 6), relative fission density (Reference 7) and relative neutron importance (Reference 7) were measured, but are not evaluated. Information for the evaluation was compiled from References 1 through 7, the experimental logbooks 8 and 9 ; additional drawings and notes provided by the experimenter; and communication with the lead experimenter, John T. Mihalczo.

  7. Orsphere: Physics Measurments For Bare, HEU(93.2)-Metal Sphere

    SciTech Connect (OSTI)

    Marshall, Margaret A.; Bess, John D.; Briggs, J. Blair; White, Christine E.; Dyrda, James P.; Tancock, Nigel P.; Mihalczo, John

    2015-03-01

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an attempt to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s (HEU-MET-FAST-001). The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. “The very accurate description of this sphere, as assembled, establishes it as an ideal benchmark for calculational methods and cross-section data files” (Reference 1). While performing the ORSphere experiments care was taken to accurately document component dimensions (±0.0001 inches), masses (±0.01 g), and material data. The experiment was also set up to minimize the amount of structural material in the sphere proximity. Two, correlated spheres were evaluated and judged to be acceptable as criticality benchmark experiments. This evaluation is given in HEU-MET-FAST-100. The second, smaller sphere was used for additional reactor physics measurements. Worth measurements (Reference 1, 2, 3 and 4), the delayed neutron fraction (Reference 3, 4 and 5) and surface material worth coefficient (Reference 1 and 2) are all measured and judged to be acceptable as benchmark data. The prompt neutron decay (Reference 6), relative fission density (Reference 7) and relative neutron importance (Reference 7) were measured, but are not evaluated. Information for the evaluation was compiled from References 1 through 7, the experimental logbooks 8 and 9 ; additional drawings and notes provided by the experimenter; and communication with the lead experimenter, John T. Mihalczo.

  8. Benchmark Specifications and Results for ßeff in a HEU Metal System Using ORSphere

    SciTech Connect (OSTI)

    Margaret A. Marshall; John D. Bess; Yevgeniy Rozhikhin

    2014-06-01

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an attempt to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s (HEU-MET-FAST-001). The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments[1]. Part of the experimental series was the measurement of the delayed neutron fraction, ßeff, using time correlation measurements and using the central void reactivity measurement. The time correlations measurements were rejected by the experimenter. The measurements using the central void reactivity measurement yielded a ßeff value of 0.00657, which agrees well with the value measured with GODIVA I (0.0066). This measurement is evaluated, found to be acceptable, and discussed in extensive detail in “ORSphere: Physics Measurements for Bare, HEU(93.2) Metal Sphere”[2]. In order to determine the delayed neutron fraction using the central void reactivity delayed neutron parameters must be used. The experimenter utilized the delayed neutron parameters set forth by Keepin, Wimment, and Zeigler[3]. If the derivation of the ßeff is repeated with different delayed neutron parameters from various modern nuclear data sets the resulting values vary greatly from the expected results.

  9. NNSA Authorizes Start-Up of Highly Enriched Uranium Materials...

    National Nuclear Security Administration (NNSA)

    Authorizes Start-Up of Highly Enriched Uranium Materials Facility at Y-12 | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the...

  10. heu | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    heu NNSA deputy administrator travels to Ukraine Earlier this month, Deputy Administrator for Defense Nuclear Nonproliferation Anne Harrington traveled to Ukraine to celebrate the 20th anniversary of the Science and Technology Center in Ukraine (STCU) and visit the Neutron Source Facility at the Kharkiv Institute of Physics and Technology. The U... NNSA keeps the promises borne out of the Nuclear Security Summit On Thursday and Friday President Obama will host his fourth and final Nuclear

  11. 2 x 2 Polyethylene Reflected and Moderated Highly Enriched Uranium System with Rhenium

    SciTech Connect (OSTI)

    A. Nichole Ellis; Jesson Hutchinson; John D. Bess; Dmitry N. Polyakov; Evgeny S. Glushkov; Alexey E. Glushkov

    2010-09-01

    The 2 2 array HEU-Re experiment was performed on the Planet universal critical assembly machine on November 4th, 2003 at the Los Alamos Critical Experiments Facility (LACEF) at Los Alamos National Laboratory (LANL). For this experiment, there were 10 units, each full unit containing four HEU foils and two rhenium foils. The top unit contained only two HEU foils and two rhenium foils. A total of 42 HEU foils were used for this experiment. Rhenium is a desirable cladding material for space nuclear power applications. This experiment consisted of HEU foils interleaved with rhenium foils and is moderated and reflected by polyethylene plates. A unit consisted of a polyethylene plate, which has a recess for rhenium foils, and four HEU foils in a single layer in the top recess of each polyethylene plate. The Planet universal criticality assembly machine has been previously used in experiments containing HEU foils interspersed with SiO2 (HEU-MET-THERM-001), Al (HEU-MET-THERM-008), MgO (HEU-MET-THERM-009), Gd foils (HEU-MET-THERM-010), 2 2 26 Al (HEU-MET-THERM-012), Fe (HEU-MET-THERM-013 and HEU-MET-THERM-015), 2 2 23 SiO2 (HEU-MET-THERM-014), 2 2 11 hastalloy plates (HEU-MET-THERM-016), and concrete (HEU-MET-THERM-018). The 2 2 array of HEU-Re is considered acceptable for use as a benchmark critical experiment.

  12. Turkey HEU Removal | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Turkey HEU Removal | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the...

  13. NMIS With Gamma Spectrometry for Attributes of Pu and HEU, Explosives and Chemical Agents

    SciTech Connect (OSTI)

    Mihalczo, J. T.; Mattingly, J. K.; Mullens, J. A.; Neal, J. S.

    2002-05-10

    The concept for the system described herein is an active/passive Nuclear Materials Identification System{sup 2} (NMIS) that incorporates gamma ray spectrometry{sup 3}. This incorporation of gamma ray spectrometry would add existing capability into this system. This Multiple Attribute System can determine a wide variety of attributes for Pu and highly enriched uranium (HEU) of which a selected subset could be chosen. This system can be built using commercial off the shelf (COTS) components. NMIS systems are at All-Russian Scientific Research Institute of Experimental Physics (VNIIEF) and Russian Federal Nuclear Center Institute of Technical Physics, (VNIITF) and measurements with Pu have been performed at VNIIEF and analyzed successfully for mass and thickness of Pu. NMIS systems are being used successfully for HEU at the Y-12 National Security Complex. The use of active gamma ray spectrometry for high explosive HE and chemical agent detection is a well known activation analysis technique, and it is incorporated here. This report describes the system, explains the attribute determination methods for fissile materials, discusses technical issues to be resolved, discusses additional development needs, presents a schedule for building from COTS components, and assembly with existing components, and discusses implementation issues such as lack of need for facility modification and low radiation exposure.

  14. Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    and Uranium-233 | Department of Energy Waste Management » Nuclear Materials & Waste » Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium and Uranium-233 Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium and Uranium-233 105-K building houses the K-Area Material Storage (KAMS) facility, designated for the consolidated storage of surplus plutonium at Savannah River Site pending disposition. The plutonium shipped to KAMS is sealed inside a

  15. Evaluation of HEU-Beryllium Benchmark Experiments to Improve Computational Analysis of Space Reactors

    SciTech Connect (OSTI)

    John D. Bess; Keith C. Bledsoe; Bradley T. Rearden

    2011-02-01

    An assessment was previously performed to evaluate modeling capabilities and quantify preliminary biases and uncertainties associated with the modeling methods and data utilized in designing a nuclear reactor such as a beryllium-reflected, highly-enriched-uranium (HEU)-O2 fission surface power (FSP) system for space nuclear power. The conclusion of the previous study was that current capabilities could preclude the necessity of a cold critical test of the FSP; however, additional testing would reduce uncertainties in the beryllium and uranium cross-section data and the overall uncertainty in the computational models. A series of critical experiments using HEU metal were performed in the 1960s and 1970s in support of criticality safety operations at the Y-12 Plant. Of the hundreds of experiments, three were identified as fast-fission configurations reflected by beryllium metal. These experiments have been evaluated as benchmarks for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). Further evaluation of the benchmark experiments was performed using the sensitivity and uncertainty analysis capabilities of SCALE 6. The data adjustment methods of SCALE 6 have been employed in the validation of an example FSP design model to reduce the uncertainty due to the beryllium cross section data.

  16. Nickel container of highly-enriched uranium bodies and sodium

    DOE Patents [OSTI]

    Zinn, Walter H.

    1976-01-01

    A fuel element comprises highly a enriched uranium bodies coated with a nonfissionable, corrosion resistant material. A plurality of these bodies are disposed in layers, with sodium filling the interstices therebetween. The entire assembly is enclosed in a fluid-tight container of nickel.

  17. Profile of World Uranium Enrichment Programs - 2007

    SciTech Connect (OSTI)

    Laughter, Mark D

    2007-11-01

    It is generally agreed that the most difficult step in building a nuclear weapon is acquiring weapons grade fissile material, either plutonium or highly enriched uranium (HEU). Plutonium is produced in a nuclear reactor, while HEU is produced using a uranium enrichment process. Enrichment is also an important step in the civil nuclear fuel cycle, in producing low enriched uranium (LEU) for use in fuel for nuclear reactors. However, the same equipment used to produce LEU for nuclear fuel can also be used to produce HEU for weapons. Safeguards at an enrichment plant are the array of assurances and verification techniques that ensure uranium is only enriched to LEU, no undeclared LEU is produced, and no uranium is enriched to HEU or secretly diverted. There are several techniques for enriching uranium. The two most prevalent are gaseous diffusion, which uses older technology and requires a lot of energy, and gas centrifuge separation, which uses more advanced technology and is more energy efficient. Gaseous diffusion plants (GDPs) provide about 40% of current world enrichment capacity, but are being phased out as newer gas centrifuge enrichment plants (GCEPs) are constructed. Estimates of current and future enrichment capacity are always approximate, due to the constant upgrades, expansions, and shutdowns occurring at enrichment plants, largely determined by economic interests. Currently, the world enrichment capacity is approximately 53 million kg-separative work units (SWU) per year, with 22 million in gaseous diffusion and 31 million in gas centrifuge plants. Another 23 million SWU/year of capacity are under construction or planned for the near future, almost entirely using gas centrifuge separation. Other less-efficient techniques have also been used in the past, including electromagnetic and aerodynamic separations, but these are considered obsolete, at least from a commercial perspective. Laser isotope separation shows promise as a possible enrichment technique of the future, but has yet to be demonstrated commercially. In the early 1980s, six countries developing gas centrifuge technology (United States, United Kingdom, Germany, the Netherlands, Japan, and Australia) along with the International Atomic Energy Agency (IAEA) and the European Atomic Energy Community (EURATOM) began developing effective safeguards techniques for GCEPs. This effort was known as the Hexapartite Safeguards Project (HSP). The HSP had the goal of maximizing safeguards effectiveness while minimizing the cost to the operator and inspectorate, and adopted several recommendations, such as the acceptance of limited-frequency unannounced access (LFUA) inspections in cascade halls, and the use of nondestructive assay (NDA) measurements and tamper-indicating seals. While only the HSP participants initially committed to implementing all the measures of the approach, it has been used as a model for the safeguards applied to GCEPs in additional states. This report provides a snapshot overview of world enrichment capacity in 2007, including profiles of the uranium enrichment programs of individual states. It is based on open-source information, which is dependent on unclassified sources and may therefore not reflect the most recent developments. In addition, it briefly describes some of the safeguards techniques being used at various enrichment plants, including implementation of HSP recommendations.

  18. GTRI's Convert Program: Minimizing the Use of Highly Enriched Uranium |

    National Nuclear Security Administration (NNSA)

    National Nuclear Security Administration GTRI's Convert Program: Minimizing the Use of Highly Enriched Uranium May 29, 2014 Mission In 2004 NNSA established the Global Threat Reduction Initiative (GTRI) in the Office of Defense Nuclear Nonproliferation to, as quickly as possible, identify, secure, remove and/or facilitate the disposition of high risk vulnerable nuclear and radiological materials around the world that pose a threat to the United States and the international community. GTRI's

  19. Initial report on characterization of excess highly enriched uranium

    SciTech Connect (OSTI)

    1996-07-01

    DOE`s Office of Fissile Materials Disposition assigned to this Y-12 division the task of preparing a report on the 174.4 metric tons of excess highly enriched U. Characterization included identification by category, gathering existing data (assay), defining the likely needed processing steps for prepping for transfer to a blending site, and developing a range of preliminary cost estimates for those steps. Focus is on making commercial reactor fuel as a final disposition path.

  20. EVALUATION OF FLOWSHEET CHANGES FOR THE HIGHLY ENRICHED URANIUM BLENDDOWN PROGRAM

    SciTech Connect (OSTI)

    Crowder, M.; Rudisill, T.; Laurinat, J.; Mickalonis, J.

    2007-10-22

    H Canyon is considering a flowsheet change for Plutonium (Pu) Contaminated Scrap (PuCS) material. The proposed change is to route dissolved PuCS material directly to a uranium (U) storage tank. As a result, the PuCS solution will bypass Head End and First U Cycle, and will be purified by solvent extraction in Second U Cycle. The PuCS solution contains appreciable amounts of boron (B) and fluoride (F{sup -}), which are currently at trace levels in the U storage tank. Though unlikely, if the B concentration in the U storage tank were to reach 1.8 g B/g U, the entire contents of the U storage tank would likely require a second pass through Second U Cycle to provide sufficient decontamination to meet the Tennessee Valley Authority (TVA) Blend Grade Highly Enriched Uranium (HEU) specification for B, which is 30 {micro}g/g U. In addition, Second U Cycle is expected to provide sufficient decontamination of F{sup -} and Pu regardless of the amount of PuCS solution sent to the storage tank. Though aluminum (Al) is not present in the PuCS solution, B can be credited as a complexant of F{sup -}. Both stability constants from the literature and Savannah River National Laboratory (SRNL) corrosion studies were documented to demonstrate that B complexation of F{sup -} in nitric acid solutions is sufficient to prevent excessive corrosion. Though B and Al complex F{sup -} to a similar degree, neither completely eliminates the presence of free F{sup -} in solution. Therefore, a limited amount of corrosion is expected even with complexed F{sup -} solutions. Tanks maintained at ambient temperature are not expected to experience significant corrosion. However, the Low Activity Waste (LAW) evaporators may be subjected to a corrosion rate of about 25 mils per year (mpy) as they reach their highest F{sup -} concentrations. The feed adjustment evaporator would only be subjected to the corrosion rate of about 25 mpy in the latter stages of the PuCS campaign. An issue that must be addressed as part of the proposed PuCS flowsheet change is that B has limited solubility in concentrated nitric acid solutions. As the proposed PuCS campaign progresses, the B concentration will increase in the U storage tank, in Second U Cycle feed, and in the 1DW stream sent to the LAW evaporators. Limitations on the B concentration in the LAW evaporators will be needed to prevent formation of boron-containing solids.

  1. The Mailbox Computer System for the IAEA verification experiment on HEU downlending at the Portsmouth Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    Aronson, A.L.; Gordon, D.M.

    2000-07-31

    IN APRIL 1996, THE UNITED STATES (US) ADDED THE PORTSMOUTH GASEOUS DIFFUSION PLANT TO THE LIST OF FACILITIES ELIGIBLE FOR THE APPLICATION OF INTERNATIONAL ATOMIC ENERGY AGENCY (IAEA) SAFEGUARDS. AT THAT TIME, THE US PROPOSED THAT THE IAEA CARRY OUT A ''VERIFICATION EXPERIMENT'' AT THE PLANT WITH RESPECT TO DOOWNBLENDING OF ABOUT 13 METRIC TONS OF HIGHLY ENRICHED URANIUM (HEU) IN THE FORM OF URANIUM HEXAFLUROIDE (UF6). DURING THE PERIOD DECEMBER 1997 THROUGH JULY 1998, THE IAEA CARRIED OUT THE REQUESTED VERIFICATION EXPERIMENT. THE VERIFICATION APPROACH USED FOR THIS EXPERIMENT INCLUDED, AMONG OTHER MEASURES, THE ENTRY OF PROCESS-OPERATIONAL DATA BY THE FACILITY OPERATOR ON A NEAR-REAL-TIME BASIS INTO A ''MAILBOX'' COMPUTER LOCATED WITHIN A TAMPER-INDICATING ENCLOSURE SEALED BY THE IAEA.

  2. D&D of the French High Enrichment Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    BEHAR, Christophe; GUIBERTEAU, Philippe; DUPERRET, Bernard; TAUZIN, Claude

    2003-02-27

    This paper describes the D&D program that is being implemented at France's High Enrichment Gaseous Diffusion Plant, which was designed to supply France's Military with Highly Enriched Uranium. This plant was definitively shut down in June 1996, following French President Jacques Chirac's decision to end production of Highly Enriched Uranium and dismantle the corresponding facilities.

  3. The Complete Burning of Weapons Grade Plutonium and Highly Enriched Uranium with (Laser Inertial Fusion-Fission Energy) LIFE Engine

    SciTech Connect (OSTI)

    Farmer, J C; Diaz de la Rubia, T; Moses, E

    2008-12-23

    The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission blanket in a fusion-fission hybrid system is subcritical, a LIFE engine can burn any fertile or fissile nuclear material, including unenriched natural or depleted U and SNF, and can extract a very high percentage of the energy content of its fuel resulting in greatly enhanced energy generation per metric ton of nuclear fuel, as well as nuclear waste forms with vastly reduced concentrations of long-lived actinides. LIFE engines could thus provide the ability to generate vast amounts of electricity while greatly reducing the actinide content of any existing or future nuclear waste and extending the availability of low cost nuclear fuels for several thousand years. LIFE also provides an attractive pathway for burning excess weapons Pu to over 99% FIMA (fission of initial metal atoms) without the need for fabricating or reprocessing mixed oxide fuels (MOX). Because of all of these advantages, LIFE engines offer a pathway toward sustainable and safe nuclear power that significantly mitigates nuclear proliferation concerns and minimizes nuclear waste. An important aspect of a LIFE engine is the fact that there is no need to extract the fission fuel from the fission blanket before it is burned to the desired final level. Except for fuel inspection and maintenance process times, the nuclear fuel is always within the core of the reactor and no weapons-attractive materials are available outside at any point in time. However, an important consideration when discussing proliferation concerns associated with any nuclear fuel cycle is the ease with which reactor fuel can be converted to weapons usable materials, not just when it is extracted as waste, but at any point in the fuel cycle. Although the nuclear fuel remains in the core of the engine until ultra deep actinide burn up is achieved, soon after start up of the engine, once the system breeds up to full power, several tons of fissile material is present in the fission blanket. However, this fissile material is widely dispersed in millions of fuel pebbles, which can be tagged as individual accountable items, and thus made difficult to divert in large quantities. This report discusses the application of the LIFE concept to nonproliferation issues, initially looking at the LIFE (Laser Inertial Fusion-Fission Energy) engine as a means of completely burning WG Pu and HEU. By combining a neutron-rich inertial fusion point source with energy-rich fission, the once-through closed fuel-cycle LIFE concept has the following characteristics: it is capable of efficiently burning excess weapons or separated civilian plutonium and highly enriched uranium; the fission blanket is sub-critical at all times (keff < 0.95); because LIFE can operate well beyond the point at which light water reactors (LWRs) need to be refueled due to burn-up of fissile material and the resulting drop in system reactivity, fuel burn-up of 99% or more appears feasible. The objective of this work is to develop LIFE technology for burning of WG-Pu and HEU.

  4. ZPR-3 Assembly 12 : A cylindrical assembly of highly enriched uranium, depleted uranium and graphite with an average {sup 235}U enrichment of 21 atom %.

    SciTech Connect (OSTI)

    Lell, R. M.; McKnight, R. D.; Perel, R. L.; Wagschal, J. J.; Nuclear Engineering Division; Racah Inst. of Physics

    2010-09-30

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 12 (ZPR-3/12) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 21 at.%. Approximately 68.9% of the total fissions in this assembly occur above 100 keV, approximately 31.1% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 9 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specifications and has historically been used as a data validation benchmark assembly. Loading of ZPR-3 Assembly 12 began in late Jan. 1958, and the Assembly 12 program ended in Feb. 1958. The core consisted of highly enriched uranium (HEU) plates, depleted uranium plates and graphite plates loaded into stainless steel drawers which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of two columns of 0.125 in.-wide (3.175 mm) HEU plates, seven columns of 0.125 in.-wide depleted uranium plates and seven columns of 0.125 in.-wide graphite plates. The length of each column was 9 in. (228.6 mm) in each half of the core. The graphite plates were included to produce a softer neutron spectrum that would be more characteristic of a large power reactor. The axial blanket consisted of 12 in. (304.8 mm) of depleted uranium behind the core. The thickness of the radial blanket was approximately 12 in. and the length of the radial blanket in each half of the matrix was 21 in. (533.4 mm). The assembly geometry approximated a right circular cylinder as closely as the square matrix tubes allowed. According to the logbook and loading records for ZPR-3/12, the reference critical configuration was loading 10 which was critical on Feb. 5, 1958. The subsequent loadings were very similar but less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/12 loading 10 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. An accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly, even a clean core such as ZPR-3/12 loading 10. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must d

  5. ZPR-3 Assembly 6F : A spherical assembly of highly enriched uranium, depleted uranium, aluminum and steel with an average {sup 235}U enrichment of 47 atom %.

    SciTech Connect (OSTI)

    Lell, R. M.; McKnight, R. D; Schaefer, R. W.; Nuclear Engineering Division

    2010-09-30

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 6 consisted of six phases, A through F. In each phase a critical configuration was constructed to simulate a very simple shape such as a slab, cylinder or sphere that could be analyzed with the limited analytical tools available in the 1950s. In each case the configuration consisted of a core region of metal plates surrounded by a thick depleted uranium metal reflector. The average compositions of the core configurations were essentially identical in phases A - F. ZPR-3 Assembly 6F (ZPR-3/6F), the final phase of the Assembly 6 program, simulated a spherical core with a thick depleted uranium reflector. ZPR-3/6F was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 47 at.%. Approximately 81.4% of the total fissions in this assembly occur above 100 keV, approximately 18.6% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 7 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specifications and has historically been used as a data validation benchmark assembly. Loading of ZPR-3/6F began in late December 1956, and the experimental measurements were performed in January 1957. The core consisted of highly enriched uranium (HEU) plates, depleted uranium plates, perforated aluminum plates and stainless steel plates loaded into aluminum drawers, which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of three columns of 0.125 in.-wide (3.175 mm) HEU plates, three columns of 0.125 in.-wide depleted uranium plates, nine columns of 0.125 in.-wide perforated aluminum plates and one column of stainless steel plates. The maximum length of each column of core material in a drawer was 9 in. (228.6 mm). Because of the goal to produce an approximately spherical core, core fuel and diluent column lengths generally varied between adjacent drawers and frequently within an individual drawer. The axial reflector consisted of depleted uranium plates and blocks loaded in the available space in the front (core) drawers, with the remainder loaded into back drawers behind the front drawers. The radial reflector consisted of blocks of depleted uranium loaded directly into the matrix tubes. The assembly geometry approximated a reflected sphere as closely as the square matrix tubes, the drawers and the shapes of fuel and diluent plates allowed. According to the logbook and loading records for ZPR-3/6F, the reference critical configuration was loading 5 which was critical on January 4, 1957. The subsequent loadings were very similar but were less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/6F loading 5 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. A very accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly. This is especially true of ZPR-3/6F because of the complex core loading required to approximate a sphere with rectangular plates in a square matrix.

  6. Two U.S. University Research Reactors to be Converted From Highly Enriched

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Uranium to Low-Enriched Uranium | Department of Energy U.S. University Research Reactors to be Converted From Highly Enriched Uranium to Low-Enriched Uranium Two U.S. University Research Reactors to be Converted From Highly Enriched Uranium to Low-Enriched Uranium April 11, 2005 - 11:34am Addthis WASHINGTON, D.C. - As part of the Bush administration's aggressive effort to reduce the amount of weapons-grade nuclear material worldwide, Secretary of Energy Samuel W. Bodman announced today that

  7. Profile of World Uranium Enrichment Programs-2009

    SciTech Connect (OSTI)

    Laughter, Mark D

    2009-04-01

    It is generally agreed that the most difficult step in building a nuclear weapon is acquiring fissile material, either plutonium or highly enriched uranium (HEU). Plutonium is produced in a nuclear reactor, whereas HEU is produced using a uranium enrichment process. Enrichment is also an important step in the civil nuclear fuel cycle, in producing low enriched uranium (LEU) for use as fuel for nuclear reactors to generate electricity. However, the same equipment used to produce LEU for nuclear reactor fuel can also be used to produce HEU for weapons. Safeguards at an enrichment plant are the array of assurances and verification techniques that ensure uranium is not diverted or enriched to HEU. There are several techniques for enriching uranium. The two most prevalent are gaseous diffusion, which uses older technology and requires a lot of energy, and gas centrifuge separation, which uses more advanced technology and is more energy efficient. Gaseous diffusion plants (GDPs) provide about 40% of current world enrichment capacity but are being phased out as newer gas centrifuge enrichment plants (GCEPs) are constructed. Estimates of current and future enrichment capacity are always approximate, due to the constant upgrades, expansions, and shutdowns occurring at enrichment plants, largely determined by economic interests. Currently, the world enrichment capacity is approximately 56 million kilogram separative work units (SWU) per year, with 22.5 million in gaseous diffusion and more than 33 million in gas centrifuge plants. Another 34 million SWU/year of capacity is under construction or planned for the near future, almost entirely using gas centrifuge separation. Other less-efficient techniques have also been used in the past, including electromagnetic and aerodynamic separations, but these are considered obsolete, at least from a commercial perspective. Laser isotope separation shows promise as a possible enrichment technique of the future but has yet to be demonstrated commercially. In the early 1980s, six countries developing gas centrifuge technology (United States, United Kingdom, Germany, the Netherlands, Japan, and Australia) along with the International Atomic Energy Agency and the European Atomic Energy Community began developing effective safeguards techniques for GCEPs. This effort was known as the Hexapartite Safeguards Project (HSP). The HSP had the goal of maximizing safeguards effectiveness while minimizing the cost to the operator and inspectorate, and adopted several recommendations, such as the acceptance of limited-frequency unannounced access inspections in cascade halls, and the use of nondestructive assay measurements and tamper-indicating seals. While only the HSP participants initially committed to implementing all the measures of the approach, it has been used as a model for the safeguards applied to GCEPs in additional states. Uranium enrichment capacity has continued to expand on all fronts in the last few years. GCEP capacity is expanding in anticipation of the eventual shutdown of the less-efficient GDPs, the termination of the U.S.-Russia HEU blend-down program slated for 2013, and the possible resurgence of nuclear reactor construction as part of an expected 'Nuclear Renaissance'. Overall, a clear trend in the world profile of uranium enrichment plant operation is the continued movement towards multinational projects driven by commercial and economic interests. Along this vein, the safeguards community is continuing to develop new safeguards techniques and technologies that are not overly burdensome to enrichment plant operators while delivering more effective and efficient results. This report provides a snapshot overview of world enrichment capacity in 2009, including profiles of the uranium enrichment programs of individual states. It is a revision of a 2007 report on the same topic; significant changes in world enrichment programs between the previous and current reports are emphasized. It is based entirely on open-source information, which is dependent on published sources and may therefore not be completely accurate or reflect the most recent developments. Consequently, readers should not assume that information cited here has the endorsement of either ORNL or the U.S. Department of Energy. We are merely reporting what's been reported. In addition, this report briefly describes some of the safeguards techniques being used at various enrichment plants, including implementation of HSP recommendations.

  8. NNSA Removes U.S.-Origin HEU from Jamaica, Makes the Caribbean...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Removes U.S.-Origin HEU from Jamaica, Makes the Caribbean HEU Free | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the...

  9. High-solids enrichment of thermophilic microbial communities and their enzymes on bioenergy feedstocks

    SciTech Connect (OSTI)

    Reddy, A. P.; Allgaier, M.; Singer, S.W.; Hazen, T.C.; Simmons, B.A.; Hugenholtz, P.; VanderGheynst, J.S.

    2011-04-01

    Thermophilic microbial communities that are active in a high-solids environment offer great potential for the discovery of industrially relevant enzymes that efficiently deconstruct bioenergy feedstocks. In this study, finished green waste compost was used as an inoculum source to enrich microbial communities and associated enzymes that hydrolyze cellulose and hemicellulose during thermophilic high-solids fermentation of the bioenergy feedstocks switchgrass and corn stover. Methods involving the disruption of enzyme and plant cell wall polysaccharide interactions were developed to recover xylanase and endoglucanase activity from deconstructed solids. Xylanase and endoglucanase activity increased by more than a factor of 5, upon four successive enrichments on switchgrass. Overall, the changes for switchgrass were more pronounced than for corn stover; solids reduction between the first and second enrichments increased by a factor of four for switchgrass while solids reduction remained relatively constant for corn stover. Amplicon pyrosequencing analysis of small-subunit ribosomal RNA genes recovered from enriched samples indicated rapid changes in the microbial communities between the first and second enrichment with the simplified communities achieved by the third enrichment. The results demonstrate a successful approach for enrichment of unique microbial communities and enzymes active in a thermophilic high-solids environment.

  10. CONCEPTUAL PROCESS DESCRIPTION FOR THE MANUFACTURE OF LOW-ENRICHED URANIUM-MOLYBDENUM FUEL

    SciTech Connect (OSTI)

    Daniel M. Wachs; Curtis R. Clark; Randall J. Dunavant

    2008-02-01

    The National Nuclear Security Agency Global Threat Reduction Initiative (GTRI) is tasked with minimizing the use of high-enriched uranium (HEU) worldwide. A key component of that effort is the conversion of research reactors from HEU to low-enriched uranium (LEU) fuels. The GTRI Convert Fuel Development program, previously known as the Reduced Enrichment for Research and Test Reactors program was initiated in 1978 by the United States Department of Energy to develop the nuclear fuels necessary to enable these conversions. The program cooperates with the research reactors’ operators to achieve this goal of HEU to LEU conversion without reduction in reactor performance. The programmatic mandate is to complete the conversion of all civilian domestic research reactors by 2014. These reactors include the five domestic high-performance research reactors (HPRR), namely: the High Flux Isotope Reactor at the Oak Ridge National Laboratory, the Advanced Test Reactor at the Idaho National Laboratory, the National Bureau of Standards Reactor at the National Institute of Standards and Technology, the Missouri University Research Reactor at the University of Missouri–Columbia, and the MIT Reactor-II at the Massachusetts Institute of Technology. Characteristics for each of the HPRRs are given in Appendix A. The GTRI Convert Fuel Development program is currently engaged in the development of a novel nuclear fuel that will enable these conversions. The fuel design is based on a monolithic fuel meat (made from a uranium-molybdenum alloy) clad in Al-6061 that has shown excellent performance in irradiation testing. The unique aspects of the fuel design, however, necessitate the development and implementation of new fabrication techniques and, thus, establishment of the infrastructure to ensure adequate fuel fabrication capability. A conceptual fabrication process description and rough estimates of the total facility throughput are described in this document as a basis for establishing preconceptual fabrication facility designs.

  11. A National Tracking Center for Monitoring Shipments of HEU, MOX, and Spent Nuclear Fuel: How do we implement?

    SciTech Connect (OSTI)

    Mark Schanfein

    2009-07-01

    Nuclear material safeguards specialists and instrument developers at US Department of Energy (USDOE) National Laboratories in the United States, sponsored by the National Nuclear Security Administration (NNSA) Office of NA-24, have been developing devices to monitor shipments of UF6 cylinders and other radioactive materials , . Tracking devices are being developed that are capable of monitoring shipments of valuable radioactive materials in real time, using the Global Positioning System (GPS). We envision that such devices will be extremely useful, if not essential, for monitoring the shipment of these important cargoes of nuclear material, including highly-enriched uranium (HEU), mixed plutonium/uranium oxide (MOX), spent nuclear fuel, and, potentially, other large radioactive sources. To ensure nuclear material security and safeguards, it is extremely important to track these materials because they contain so-called “direct-use material” which is material that if diverted and processed could potentially be used to develop clandestine nuclear weapons . Large sources could be used for a dirty bomb also known as a radioactive dispersal device (RDD). For that matter, any interdiction by an adversary regardless of intent demands a rapid response. To make the fullest use of such tracking devices, we propose a National Tracking Center. This paper describes what the attributes of such a center would be and how it could ultimately be the prototype for an International Tracking Center, possibly to be based in Vienna, at the International Atomic Energy Agency (IAEA).

  12. Nuclear Design of Fissile Pu and HEU LIFE Engine - NA22 (Technical...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: Nuclear Design of Fissile Pu and HEU LIFE Engine - NA22 Citation Details In-Document Search Title: Nuclear Design of Fissile Pu and HEU LIFE Engine - NA22 ...

  13. Nuclear Design of Fissile Pu and HEU LIFE Engine - NA22 (Technical...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: Nuclear Design of Fissile Pu and HEU LIFE Engine - NA22 Citation Details In-Document Search Title: Nuclear Design of Fissile Pu and HEU LIFE Engine - NA22 You ...

  14. Utilization of non-weapons-grade plutonium and highly enriched uranium with

    Office of Scientific and Technical Information (OSTI)

    breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water (Journal Article) | SciTech Connect Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water Citation Details In-Document Search Title: Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water A

  15. DOE to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Weapons Stockpile | Department of Energy to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear Weapons Stockpile DOE to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear Weapons Stockpile November 7, 2005 - 12:38pm Addthis Will Be Redirected to Naval Reactors, Down-blended or Used for Space Programs WASHINGTON, DC - Secretary of Energy Samuel W. Bodman today announced that the Department of Energy's (DOE) National Nuclear Security Administration (NNSA) will

  16. ZPR-3 Assembly 11 : A cylindrical sssembly of highly enriched uranium and depleted uranium with an average {sup 235}U enrichment of 12 atom % and a depleted uranium reflector.

    SciTech Connect (OSTI)

    Lell, R. M.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; National Security; Inst. of Physics and Power Engineering

    2010-09-30

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specificationsa and has historically been used as a data validation benchmark assembly. Loading of ZPR-3 Assembly 11 began in early January 1958, and the Assembly 11 program ended in late January 1958. The core consisted of highly enriched uranium (HEU) plates and depleted uranium plates loaded into stainless steel drawers, which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of two columns of 0.125 in.-wide (3.175 mm) HEU plates, six columns of 0.125 in.-wide (3.175 mm) depleted uranium plates and one column of 1.0 in.-wide (25.4 mm) depleted uranium plates. The length of each column was 10 in. (254.0 mm) in each half of the core. The axial blanket consisted of 12 in. (304.8 mm) of depleted uranium behind the core. The thickness of the depleted uranium radial blanket was approximately 14 in. (355.6 mm), and the length of the radial blanket in each half of the matrix was 22 in. (558.8 mm). The assembly geometry approximated a right circular cylinder as closely as the square matrix tubes allowed. According to the logbook and loading records for ZPR-3/11, the reference critical configuration was loading 10 which was critical on January 21, 1958. Subsequent loadings were very similar but less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/11 loading 10 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. A very accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly, even a clean core such as ZPR-3/11 loading 10. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must do this without increasing the total uncertain

  17. SPERT Destructive Test - I on Aluminum, Highly Enriched Plate Type Core

    SciTech Connect (OSTI)

    2011-04-05

    SPERT - Special Power Excursion Reactor Tests Destructive Test number 1 On Aluminum, Highly Enriched Plate Type Core. A test studying the behavior of the reactor under destructive conditions on a light water moderated pool-type reactor with a plate-type core.

  18. SPERT Destructive Test - I on Aluminum, Highly Enriched Plate Type Core

    ScienceCinema (OSTI)

    None

    2014-05-07

    SPERT - Special Power Excursion Reactor Tests Destructive Test number 1 On Aluminum, Highly Enriched Plate Type Core. A test studying the behavior of the reactor under destructive conditions on a light water moderated pool-type reactor with a plate-type core.

  19. Isotope Enrichment Detection by Laser Ablation - Dual Tunable Diode Laser Absorption Spectrometry

    SciTech Connect (OSTI)

    Anheier, Norman C.; Bushaw, Bruce A.

    2009-07-01

    The rapid global expansion of nuclear energy is motivating the expedited development of new safeguards technology to mitigate potential proliferation threats arising from monitoring gaps within the uranium enrichment process. Current onsite enrichment level monitoring methods are limited by poor sensitivity and accuracy performance. Offsite analysis has better performance, but this approach requires onsite hand sampling followed by time-consuming and costly post analysis. These limitations make it extremely difficult to implement comprehensive safeguards accounting measures that can effectively counter enrichment facility misuse. In addition, uranium enrichment by modern centrifugation leads to a significant proliferation threat, since the centrifuge cascades can quickly produce a significant quantity of highly enriched uranium (HEU). The Pacific Northwest National Laboratory is developing an engineered safeguards approach having continuous aerosol particulate collection and uranium isotope analysis to provide timely detection of HEU production in a low enriched uranium facility. This approach is based on laser vaporization of aerosol particulate samples, followed by wavelength tuned laser diode spectroscopy, to characterize the 235U/238U isotopic ratio by subtle differences in atomic absorption wavelengths arising from differences in each isotopes nuclear mass, volume, and spin (hyperfine structure for 235U). Environmental sampling media is introduced into a small, reduced pressure chamber, where a focused pulsed laser vaporizes a 10 to 20-m sample diameter. The ejected plasma forms a plume of atomic vapor. A plume for a sample containing uranium has atoms of the 235U and 238U isotopes present. Tunable diode lasers are directed through the plume to selectively excite each isotope and their presence is detected by monitoring absorbance signals on a shot-to-shot basis. Single-shot detection sensitivity approaching the femtogram range and abundance uncertainty less than 10% have been demonstrated with measurements on surrogate materials. In this paper we present measurement results on samples containing background materials (e.g., dust, minerals, soils) laced with micron-sized target particles having isotopic ratios ranging from 1 to 50%.

  20. Performance and safety parameters for the high flux isotope reactor

    SciTech Connect (OSTI)

    Ilas, G. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831-6172 (United States); Primm III, T. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831-6172 (United States); Primm Consulting, LLC, 945 Laurel Hill Road, Knoxville, TN 37923 (United States)

    2012-07-01

    A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDF/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data. (authors)

  1. Performance and Safety Parameters for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [Primm Consulting, LLC

    2012-01-01

    A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDV/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared when available with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data.

  2. R and D of crystal scintillators from enriched isotopes for high sensitivity double ? decay experiments

    SciTech Connect (OSTI)

    Danevich, F. A. [Institute for Nuclear Research, Kyiv (Ukraine)

    2013-12-30

    Experiments to search for neutrinoless double beta decay enters to a new phase when a sensitivity on the level of T{sub 1/2}?10{sup 26}?10{sup 28} yr is required. Scintillating low temperature detectors possess important properties required for high-sensitivity double beta decay experiments: presence of elements of interest, high energy resolution and detection efficiency, low level of background thanks to excellent particle discrimination ability. High concentration of isotope of interest and as low as possible radioactive contamination are important requirements to crystal scintillators. Other crucial issues are maximal output of detectors and minimal loss of enriched materials. Prospects of several scintillation materials, enriched in isotopes promising for double beta decay experiments, are discussed.

  3. Establishing Specifications for Low Enriched Uranium Fuel Operations Conducted Outside the High Flux Isotope Reactor Site

    SciTech Connect (OSTI)

    Pinkston, Daniel [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL

    2010-10-01

    The National Nuclear Security Administration (NNSA) has funded staff at Oak Ridge National Laboratory (ORNL) to study the conversion of the High Flux Isotope Reactor (HFIR) from the current, high enriched uranium fuel to low enriched uranium fuel. The LEU fuel form is a metal alloy that has never been used in HFIR or any HFIR-like reactor. This report provides documentation of a process for the creation of a fuel specification that will meet all applicable regulations and guidelines to which UT-Battelle, LLC (UTB) the operating contractor for ORNL - must adhere. This process will allow UTB to purchase LEU fuel for HFIR and be assured of the quality of the fuel being procured.

  4. Microsoft PowerPoint - 10_BILL_WANDERER_REVISED HEU Agreement...

    National Nuclear Security Administration (NNSA)

    Russia processed HEU at four sites: Mayak Production Association (MPA) Siberian Group of Chemical Enterprises (SChE) Electrochemical Plant (ECP) Ural Electrochemical Integrated ...

  5. NNSA Awards Agreement to Establish Non-HEU-Based Production of...

    National Nuclear Security Administration (NNSA)

    reliable supply of the critical medical isotope Mo-99, produced without the use of proliferation-sensitive HEU. Today's announcement marks a significant step in these efforts. ...

  6. Modeling and Simulations for the High Flux Isotope Reactor Cycle 400

    Office of Scientific and Technical Information (OSTI)

    (Technical Report) | SciTech Connect Modeling and Simulations for the High Flux Isotope Reactor Cycle 400 Citation Details In-Document Search Title: Modeling and Simulations for the High Flux Isotope Reactor Cycle 400 A concerted effort over the past few years has been focused on enhancing the core model for the High Flux Isotope Reactor (HFIR), as part of a comprehensive study for HFIR conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel. At this time, the core

  7. Domestic U.S. Reactor Conversions: Fact Sheet | National Nuclear...

    National Nuclear Security Administration (NNSA)

    the use of highly-enriched uranium (HEU) fuel to low-enriched uranium (LEU) as a key ... the use of highly-enriched uranium (HEU) fuel to low-enriched uranium (LEU) as a key ...

  8. Draft EA for the Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    EA for the Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany DOE/EA-1977 DRAFT ENVIRONMENTAL ASSESSMENT FOR THE ACCEPTANCE AND DISPOSITION OF SPENT NUCLEAR FUEL CONTAINING U.S.-ORIGIN HIGHLY ENRICHED URANIUM FROM THE FEDERAL REPUBLIC OF GERMANY January 2016 U.S. DEPARTMENT OF ENERGY SAVANNAH RIVER OPERATIONS OFFICE AIKEN, SOUTH CAROLINA Draft EA for the Acceptance and Disposition of Spent Nuclear Fuel Containing

  9. Defining the needs for gas centrifuge enrichment plants advanced safeguards

    SciTech Connect (OSTI)

    Boyer, Brian David; Erpenbeck, Heather H; Miller, Karen A; Swinhoe, Martyn T; Ianakiev, Kiril; Marlow, Johnna B

    2010-04-05

    Current safeguards approaches used by the International Atomic Energy Agency (IAEA) at gas centrifuge enrichment plants (GCEPs) need enhancement in order to verify declared low-enriched (LEU) production, detect undeclared LEU production and detect highly enriched uranium (HEU) production with adequate detection probability using nondestructive assay (NDA) techniques. At present inspectors use attended systems, systems needing the presence of an inspector for operation, during inspections to verify the mass and {sup 235}U enrichment of declared UF{sub 6} containers used in the process of enrichment at GCEPs. In verifying declared LEU production, the inspectors also take samples for off-site destructive assay (DA) which provide accurate data, with 0.1% to 0.5% measurement uncertainty, on the enrichment of the UF{sub 6} feed, tails, and product. However, taking samples of UF{sub 6} for off-site analysis is a much more labor and resource intensive exercise for the operator and inspector. Furthermore, the operator must ship the samples off-site to the IAEA laboratory which delays the timeliness of results and interruptions to the continuity of knowledge (CofK) of the samples during their storage and transit. This paper contains an analysis of possible improvements in unattended and attended NDA systems such as process monitoring and possible on-site analysis of DA samples that could reduce the uncertainty of the inspector's measurements and provide more effective and efficient IAEA GCEPs safeguards. We also introduce examples advanced safeguards systems that could be assembled for unattended operation.

  10. Fuel Grading Study on a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Ilas, Germina; Primm, Trent

    2009-11-01

    An engineering design study that would enable the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium fuel is ongoing at Oak Ridge National Laboratory. The computational models used to search for a low-enriched uranium (LEU) fuel design that would meet the requirements for the conversion study, and the recent results obtained with these models during FY 2009, are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating high-enriched uranium fuel core. These studies indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations.

  11. Compact reaction cell for homogenizing and down-blanding highly enriched uranium metal

    DOE Patents [OSTI]

    McLean, II, William; Miller, Philip E.; Horton, James A.

    1995-01-01

    The invention is a specialized reaction cell for converting uranium metal to uranium oxide. In a preferred form, the reaction cell comprises a reaction chamber with increasing diameter along its length (e.g. a cylindrical chamber having a diameter of about 2 inches in a lower portion and having a diameter of from about 4 to about 12 inches in an upper portion). Such dimensions are important to achieve the necessary conversion while at the same time affording criticality control and transportability of the cell and product. The reaction chamber further comprises an upper port and a lower port, the lower port allowing for the entry of reactant gasses into the reaction chamber, the upper port allowing for the exit of gasses from the reaction chamber. A diffuser plate is attached to the lower port of the reaction chamber and serves to shape the flow of gas into the reaction chamber. The reaction cell further comprises means for introducing gasses into the reaction chamber and a heating means capable of heating the contents of the reaction chamber. The present invention also relates to a method for converting uranium metal to uranium oxide in the reaction cell of the present invention. The invention is useful for down-blending highly enriched uranium metal by the simultaneous conversion of highly enriched uranium metal and natural or depleted uranium metal to uranium oxide within the reaction cell.

  12. Compact reaction cell for homogenizing and down-blending highly enriched uranium metal

    DOE Patents [OSTI]

    McLean, W. II; Miller, P.E.; Horton, J.A.

    1995-05-02

    The invention is a specialized reaction cell for converting uranium metal to uranium oxide. In a preferred form, the reaction cell comprises a reaction chamber with increasing diameter along its length (e.g. a cylindrical chamber having a diameter of about 2 inches in a lower portion and having a diameter of from about 4 to about 12 inches in an upper portion). Such dimensions are important to achieve the necessary conversion while at the same time affording criticality control and transportability of the cell and product. The reaction chamber further comprises an upper port and a lower port, the lower port allowing for the entry of reactant gases into the reaction chamber, the upper port allowing for the exit of gases from the reaction chamber. A diffuser plate is attached to the lower port of the reaction chamber and serves to shape the flow of gas into the reaction chamber. The reaction cell further comprises means for introducing gases into the reaction chamber and a heating means capable of heating the contents of the reaction chamber. The present invention also relates to a method for converting uranium metal to uranium oxide in the reaction cell of the present invention. The invention is useful for down-blending highly enriched uranium metal by the simultaneous conversion of highly enriched uranium metal and natural or depleted uranium metal to uranium oxide within the reaction cell. 4 figs.

  13. Transition Core Properties during Conversion of the NBSR from HEU to LEU Fuel

    SciTech Connect (OSTI)

    Hanson A. L.; Diamond D.

    2013-10-31

    The transition of the NBSR from HEU to LEU fuel is challenging due to reactivity constraints and the need to maintain an uninterrupted science program, the mission of the NBSR. The transition cannot occur with a full change of HEU to LEU fuel elements since the excess reactivity would be large enough that the NBSR would violate the technical specification for shutdown margin. Manufacturing LEU fuel elements to represent irradiated fuel elements would be cost prohibitive since 26 one-of-a-kind fuel elements would need to be manufactured. For this report a gradual transition from the present HEU fuel to the proposed LEU fuel was studied. The gradual change approach would follow the present fuel management scheme and replace four HEU fuel elements with four LEU fuel elements each cycle. This manuscript reports the results of a series of calculations to predict the neutronic characteristics and how the neutronics will change during the transition from HEU to LEU in the NBSR.

  14. Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the Government of the United States and the Government of the Russian Federation has on the

    Energy Savers [EERE]

    Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the Government of the United States of America and the Government of the Russian Federation has on the Domestic Uranium Mining, Conversion, and Enrichment Industries and the Operation of the Gaseous Diffusion Plant 2008 Information Date: December 31, 2008 1 Introduction The Agreement Between the Government of the United States of America and the Government of the Russian Federation

  15. Analysis of the effectiveness of gas centrifuge enrichment plants advanced safeguards

    SciTech Connect (OSTI)

    Boyer, Brian David; Erpenbeck, Heather H; Miller, Karen A; Swinjoe, Martyn T; Ianakiev, Kiril D; Marlow, Johnna B

    2010-01-01

    Current safeguards approaches used by the International Atomic Energy Agency (IAEA) at gas centrifuge enrichment plants (GCEPs) need enhancement in order to verify declared low-enriched uranium (LEU) production, detect undeclared LEU production and detect highly enriched uranium (HEU) production with adequate detection probability using non destructive assay (NDA) techniques. At present inspectors use attended systems, systems needing the presence of an inspector for operation, during inspections to verify the mass and 235U enrichment of declared UF6 containers used in the process of enrichment at GCEPs. This paper contains an analysis of possible improvements in unattended and attended NDA systems including process monitoring and possible on-site destructive assay (DA) of samples that could reduce the uncertainty of the inspector's measurements. These improvements could reduce the difference between the operator's and inspector's measurements providing more effective and efficient IAEA GCEPs safeguards. We also explore how a few advanced safeguards systems could be assembled for unattended operation. The analysis will focus on how unannounced inspections (UIs), and the concept of information-driven inspections (IDS) can affect probability of detection of the diversion of nuclear materials when coupled to new GCEPs safeguards regimes augmented with unattended systems.

  16. NNSA Awards Funding to Accelerate Non-HEU-Based Production of...

    National Nuclear Security Administration (NNSA)

    objectives of eliminating the use of proliferation-sensitive highly enriched uranium ... to reduce the risk of nuclear proliferation while providing stability to an ...

  17. Consideration of critically when directly disposing highly enriched spent nuclear fuel in unsaturated tuff: Bounding estimates

    SciTech Connect (OSTI)

    Rechard, R.P.; Tierney, M.S.; Sanchez, L.C.; Martell, M.-A.

    1996-05-01

    This report presents one of 2 approaches (bounding calculations) which were used in a 1994 study to examine the possibility of a criticality in a repository. Bounding probabilities, although rough, point to the difficulty of creating conditions under which a critical mass could be assembled (container corrosion, separation of neutron absorbers from fissile material, collapse or precipitation of fissile material) and how significant the geochemical and hydrologic phenomena are. The study could not conceive of a mechanism consistent with conditions under which an atomic explosion could occur. Should a criticality occur in or near a container in the future, boundary consequence calculations showed that fissions from one critical event (<10{sup 20} fissions, if similar to aqueous and metal accidents and experiments) are quite small compared to the amount of fissions represented by the spent fuel itself. If it is assumed that the containers necessary to hold the highly enriched spent fuel went critical once per day for 1 million years, creating an energy release of about 10{sup 20} fissions, the number of fissions equals about 10{sup 28}, which corresponds to only 1% of the fission inventory in a repository containing 70,000 metric tons of heavy metal, the expected size for the proposed repository at Yucca Mountain, Nevada.

  18. Field Trial of LANL On-Line Advanced Enrichment Monitor for UF6 GCEP

    SciTech Connect (OSTI)

    Ianakiev, Kiril D.; Lombardi, Marcie; MacArthur, Duncan W.; Parker, Robert F.; Smith, Morag K.; Keller, Clifford; Friend, Peter; Dunford, Andrew

    2012-07-13

    The outline of this presentation is: (1) Technology basis of on-line enrichment monitoring; (2) Timescale of trial; (3) Description of installed equipment; (4) Photographs; (5) Results; (6) Possible further development; and (7) Conclusions. Summary of the good things about the Advanced Enrichment Monitor (AEM) performance is: (1) High accuracy - normally better than 1% relative, (2) Active system as accurate as passive system, (3) Fast and accurate detection of enrichment changes, (4) Physics is well understood, (5) Elegant method for capturing pressure signal, and (6) Data capture is automatic, low cost and fast. A couple of negative things are: (1) Some jumps in measured passive enrichment - of around +2% relative (due to clock errors?); and (2) Data handling and evaluation is off-line, expensive and very slow. Conclusions are: (1) LANL AEM is being tested on E23 plant at Capenhurst; (2) The trial is going very well; (3) AEM could detect production of HEU at potentially much lower cost than existing CEMO; (4) AEM can measure {sup 235}U assay accurately; (5) Active system using X-Ray source would avoid need for pressure measurement; (6) Substantial work lies ahead to go from current prototype to a production instrument.

  19. Development of the laser isotope separation method (AVLIS) for obtaining weight amounts of highly enriched {sup 150}Nd isotope

    SciTech Connect (OSTI)

    Babichev, A P; Grigoriev, Igor' S; Grigoriev, A I; Dorovskii, A P; D'yachkov, Aleksei B; Kovalevich, S K; Kochetov, V A; Kuznetsov, V A; Labozin, Valerii P; Matrakhov, A V; Mironov, Sergei M; Nikulin, Sergei A; Pesnya, A V; Timofeev, N I; Firsov, Valerii A; Tsvetkov, G O; Shatalova, G G

    2005-10-31

    Results obtained at the first stage of development of the experimental technique for obtaining weight amounts of the highly enriched {sup 150}Nd isotope by laser photoionisation are presented. The vaporiser and the laser are designed, and various methods of irradiation of neodymium vapour and extraction of photoions are tested. The product yield {approx}40 mg h{sup -1} for the {approx}60% enrichment and 25 mg h{sup -1} for the {approx}65% enrichment is achieved for a vaporiser of length 27 cm. The cost of constructing the facility for preparing 50 kg of the {sup 150}Nd isotope, intended for determining the neutrino mass, is estimated. This estimate shows that the cost of production can be lowered by a factor of 5-7 compared to the electromagnetic method. (invited paper)

  20. US Removes Last Remaining HEU from Czech Republic, Sets Nonproliferati...

    National Nuclear Security Administration (NNSA)

    never acquire a nuclear weapon, and our work ... into low enriched uranium (LEU) for use in power reactors. ... NNSA maintains and enhances the safety, security, ...

  1. Transmutation Analysis of Enriched Uranium and Deep Burn High Temperature Reactors

    SciTech Connect (OSTI)

    Michael A. Pope

    2012-07-01

    High temperature reactors (HTRs) have been under consideration for production of electricity, process heat, and for destruction of transuranics for decades. As part of the transmutation analysis efforts within the Fuel Cycle Research and Development (FCR&D) campaign, a need was identified for detailed discharge isotopics from HTRs for use in the VISION code. A conventional HTR using enriched uranium in UCO fuel was modeled having discharge burnup of 120 GWd/MTiHM. Also, a deep burn HTR (DB-HTR) was modeled burning transuranic (TRU)-only TRU-O2 fuel to a discharge burnup of 648 GWd/MTiHM. For each of these cases, unit cell depletion calculations were performed with SCALE/TRITON. Unit cells were used to perform this analysis using SCALE 6.1. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were first set by using Serpent calculations to match a spectral index between unit cell and whole core domains. In the case of the DB-HTR, the unit cell which was arrived at in this way conserved the ratio of fuel to moderator found in a single block of fuel. In the conventional HTR case, a larger moderator-to-fuel ratio than that of a single block was needed to simulate the whole core spectrum. Discharge isotopics (for 500 nuclides) and one-group cross-sections (for 1022 nuclides) were delivered to the transmutation analysis team. This report provides documentation for these calculations. In addition to the discharge isotopics, one-group cross-sections were provided for the full list of 1022 nuclides tracked in the transmutation library.

  2. Laser and gas centrifuge enrichment

    SciTech Connect (OSTI)

    Heinonen, Olli

    2014-05-09

    Principles of uranium isotope enrichment using various laser and gas centrifuge techniques are briefly discussed. Examples on production of high enriched uranium are given. Concerns regarding the possibility of using low end technologies to produce weapons grade uranium are explained. Based on current assessments commercial enrichment services are able to cover the global needs of enriched uranium in the foreseeable future.

  3. Uranium Enrichment Standards of the Y-12 Nuclear Detection and Sensor Testing Center

    SciTech Connect (OSTI)

    Cantrell, J.

    2012-05-23

    The Y-12 National Security Complex has recently fabricated and characterized a new series of metallic uranium standards for use in the Nuclear Detection and Sensor Testing Center (NDSTC). Ten uranium metal disks with enrichments varying from 0.2 to 93.2% {sup 235}U were designed to provide researchers access to a wide variety of measurement scenarios in a single testing venue. Special care was taken in the selection of the enrichments in order to closely bracket the definitions of reactor fuel at 4% {sup 235}U and that of highly enriched uranium (HEU) at 20% {sup 235}U. Each standard is well characterized using analytical chemistry as well as a series of gamma-ray spectrometry measurements. Gamma-ray spectra of these standards are being archived in a reference library for use by customers of the NDSTC. A software database tool has been created that allows for easier access and comparison of various spectra. Information provided through the database includes: raw count data (including background spectra), regions of interest (ROIs), and full width half maximum calculations. Input is being sought from the user community on future needs including enhancements to the spectral database and additional Uranium standards, shielding configurations and detector types. A related presentation are planned for the INMM 53rd Annual Meeting (Hull, et al.), which describe new uranium chemical compound standards and testing opportunities at Y-12 Nuclear Detection and Sensor Testing Center (NDSTC).

  4. Uranium Mining, Conversion, and Enrichment Industries

    Office of Environmental Management (EM)

    ... Down-blending of HEU NNSA contracts with WesDyne International for down-blending of HEU to ... Under the terms of the contract with WesDyne, DOE can use a mix of money and ...

  5. Production of highly-enriched 134Ba for a reference material for isotope dilution mass spectrometry measurements

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Horkley, J. J.; Carney, K. P.; Gantz, E. M.; Davies, J. E.; Lewis, R. R.; Crow, J. P.; Poole, C. A.; Grimes, T. S.; Giglio, J. J.

    2015-03-17

    Isotope dilution mass spectrometry (IDMS) is an analytical technique capable of providing accurate and precise quantitation of trace isotope abundance and assay providing measurement uncertainties below 1 %. To achieve these low uncertainties, the IDMS method ideally utilizes chemically pure “spike” solutions that consist of a single highly enriched isotope that is well-characterized relating to the abundance of companion isotopes and concentration in solution. To address a current demand for accurate 137Cs/137Ba ratio measurements for “age” determination of radioactive 137Cs sources, Idaho National Laboratory (INL) is producing enriched 134Ba isotopes that are tobe used for IDMS spikes to accurately determinemore » 137Ba accumulation from the decay of 137Cs. The final objective of this work it to provide a homogenous set of reference materials that the National Institute of Standards and Technology can certify as standard reference materials used for IDMS. The process that was developed at INL for the separation and isolation of Ba isotopes, chemical purification of the isotopes in solution, and the encapsulation of the materials will be described.« less

  6. Production of highly-enriched 134Ba for a reference material for isotope dilution mass spectrometry measurements

    SciTech Connect (OSTI)

    Horkley, J. J.; Carney, K. P.; Gantz, E. M.; Davies, J. E.; Lewis, R. R.; Crow, J. P.; Poole, C. A.; Grimes, T. S.; Giglio, J. J.

    2015-03-17

    Isotope dilution mass spectrometry (IDMS) is an analytical technique capable of providing accurate and precise quantitation of trace isotope abundance and assay providing measurement uncertainties below 1 %. To achieve these low uncertainties, the IDMS method ideally utilizes chemically pure “spike” solutions that consist of a single highly enriched isotope that is well-characterized relating to the abundance of companion isotopes and concentration in solution. To address a current demand for accurate 137Cs/137Ba ratio measurements for “age” determination of radioactive 137Cs sources, Idaho National Laboratory (INL) is producing enriched 134Ba isotopes that are tobe used for IDMS spikes to accurately determine 137Ba accumulation from the decay of 137Cs. The final objective of this work it to provide a homogenous set of reference materials that the National Institute of Standards and Technology can certify as standard reference materials used for IDMS. The process that was developed at INL for the separation and isolation of Ba isotopes, chemical purification of the isotopes in solution, and the encapsulation of the materials will be described.

  7. Production of highly-enriched 134Ba for a reference material for isotope dilution mass spectrometry measurements

    SciTech Connect (OSTI)

    J.J. Horkley; K.P E.M. Gantz; J.E. Davis; R.R. Lewis; J.P. Crow; C.A. Poole; T.S. Grimes; J.J. Giglio

    2015-03-01

    t Isotope dilution mass spectrometry (IDMS) is an analytical technique capable of providing accurate and precise quantitation of trace isotope abundance and assay providing measurement uncertainties below 1 %. To achieve these low uncertainties, the IDMS method ideally utilizes chemically pure spike solutions that consist of a single highly enriched isotope that is well-characterized relating to the abundance of companion isotopes and concentration in solution. To address a current demand for accurate 137Cs/137Ba ratio measurements for age determination of radioactive 137Cs sources, Idaho National Laboratory (INL) is producing enriched 134Ba isotopes that are tobe used for IDMS spikes to accurately determine 137Ba accumulation from the decay of 137Cs. The final objective of this work it to provide a homogenous set of reference materials that the National Institute of Standards and Technology can certify as standard reference materials used for IDMS. The process that was developed at INL for the separation and isolation of Ba isotopes, chemical purification of the isotopes in solution,

  8. Production of highly-enriched 134Ba for a reference material for isotope dilution mass spectrometry measurements

    SciTech Connect (OSTI)

    Horkley, J. J.; Carney, K. P.; Gantz, E. M.; Davies, J. E.; Lewis, R. R.; Crow, J. P.; Poole, C. A.; Grimes, T. S.; Giglio, J. J.

    2015-03-17

    Isotope dilution mass spectrometry (IDMS) is an analytical technique capable of providing accurate and precise quantitation of trace isotope abundance and assay providing measurement uncertainties below 1 %. To achieve these low uncertainties, the IDMS method ideally utilizes chemically pure spike solutions that consist of a single highly enriched isotope that is well-characterized relating to the abundance of companion isotopes and concentration in solution. To address a current demand for accurate 137Cs/137Ba ratio measurements for age determination of radioactive 137Cs sources, Idaho National Laboratory (INL) is producing enriched 134Ba isotopes that are tobe used for IDMS spikes to accurately determine 137Ba accumulation from the decay of 137Cs. The final objective of this work it to provide a homogenous set of reference materials that the National Institute of Standards and Technology can certify as standard reference materials used for IDMS. The process that was developed at INL for the separation and isolation of Ba isotopes, chemical purification of the isotopes in solution, and the encapsulation of the materials will be described.

  9. A Robust and Flexible Design for GCEP Unattended Online Enrichment Monitoring: An OLEM Collection Node Network

    SciTech Connect (OSTI)

    Younkin, James R; March-Leuba, Jose A; Garner, James R

    2013-01-01

    Oak Ridge National Laoratory (ORNL) has engineered an on-line enrichment monitor (OLEM) to continuously measure U-235 emissions from the UF6 gas flowing through a unit header pipe of a gas centrifuge enrichment plant (GCEP) as a component of the International Atomic Energy Agency s (IAEA) new generation of technology to support enrichment plant safeguards1. In contrast to other enrichment monitoring approaches, OLEM calibrates and corrects for the pressure and temperature dependent UF6 gas-density without external radiation sources by using the inherent unit header pipe pressure dynamics and combining U-235 gamma-ray spectrometery using a shielded NaI detector with gas pressure and temperature data near the spectrum measurement point to obtain the enrichment of the gas as a function of time. From a safeguards perspective, OLEM can provide early detection of a GCEP being misused for production of highly enriched uranium, but would not detect directly the isolation and use of a cascade within the production unit to produce HEU. OLEM may also reduce the number of samples collected for destructive assay and, if coupled with load cell monitoring, could support isotope mass balance verification and unattended cylinder verification. The earlier paper presented OLEM as one component along with shared load cells and unattended cylinder verification, in the IAEA emering toolbox for unattended instruments at GCEPs1 and described the OLEM concept and how previous modeling studies and field measurements helped confirm the viability of a passive on-line enrichment monitor for meeting IAEA objectives and to support the development of performance targets. Phase I of the United States Support Program (USSP) OLEM project completed a preliminary hardware, software and communications design; phase II will build and test field prototypes in controlled laboratory settings and then at an operational facility. That paper also discussed many of the OLEM collection node commercial off the shelf (COTS) components and summarized the OLEM collection node data security provisions. This paper will discuss a secure and redundant network of OLEM collection nodes, auxiliary detection units and supporting junction boxes distributed throughout a facility for monitoring enrichment on product, feed and tails unit header pipes; the purpose and capability of the built-in Electronic Optical Sealing System (EOSS) network gateway; and a network approach for obtaining reliable and authenticated pressure measurements.

  10. Uranium enrichment

    SciTech Connect (OSTI)

    Not Available

    1991-04-01

    This book presents the GAO's views on the Department of Energy's (DOE) program to develop a new uranium enrichment technology, the atomic vapor laser isotope separation process (AVLIS). Views are drawn from GAO's ongoing review of AVLIS, in which the technical, program, and market issues that need to be addressed before an AVLIS plant is built are examined.

  11. Thermal breeder fuel enrichment zoning

    DOE Patents [OSTI]

    Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

    1992-01-01

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

  12. Uranium enrichment

    SciTech Connect (OSTI)

    Not Available

    1991-08-01

    This paper reports that in 1990 the Department of Energy began a two-year project to illustrate the technical and economic feasibility of a new uranium enrichment technology-the atomic vapor laser isotope separation (AVLIS) process. GAO believes that completing the AVLIS demonstration project will provide valuable information about the technical viability and cost of building an AVLIS plant and will keep future plant construction options open. However, Congress should be aware that DOE still needs to adequately demonstrate AVLIS with full-scale equipment and develop convincing cost projects. Program activities, such as the plant-licensing process, that must be completed before a plant is built, could take many years. Further, an updated and expanded uranium enrichment analysis will be needed before any decision is made about building an AVLIS plant. GAO, which has long supported legislation that would restructure DOE's uranium enrichment program as a government corporation, encourages DOE's goal of transferring AVLIS to the corporation. This could reduce the government's financial risk and help ensure that the decision to build an AVLIS plant is based on commercial concerns. DOE, however, has no alternative plans should the government corporation not be formed. Further, by curtailing a planned public access program, which would have given private firms an opportunity to learn about the technology during the demonstration project, DOE may limit its ability to transfer AVLIS to the private sector.

  13. NNSA Completes Conversion of the Budapest Research Reactor and...

    National Nuclear Security Administration (NNSA)

    Atomic Energy Research Institute, successfully converted the Budapest Research Reactor (BRR) from the use of highly enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. ...

  14. highly enriched uranium

    National Nuclear Security Administration (NNSA)

    NorthStar Medical Radioisotopes to further develop its technology to produce Mo-99 via neutron capture, bringing the total NNSA support to this project to the maximum of 25...

  15. Verification of the MCU precision code and ROSFOND neutron data in application to the calculations of criticality of fast reactors with highly enriched uranium

    SciTech Connect (OSTI)

    Alekseev, N. I.; Kalugin, M. A.; Kulakov, A. S.; Novoseltsev, A. P.; Sergeev, G. S.; Shkarovskiy, D. A.; Yudkevich, M. S.

    2014-12-15

    Calculation of 335 critical assemblies (benchmark experiments) with the core of highly enriched uranium and reflectors of various materials is performed. The statistical analysis of the results shows that, for all 16 materials studied, the absolute value of the most probable deviation of the calculated value of K{sub eff} from the experimental one does not exceed 0.005.

  16. removal

    National Nuclear Security Administration (NNSA)

    80 pounds) of highly enriched uranium (HEU) spent fuel from the Institute of Nuclear Physics (INP) in Almaty, Kazakhstan. The HEU was transported via two air shipments to a...

  17. Pulsed Neutron Measurments With A DT Neutron Generator for an Annular HEU Uranium Metal Casting

    SciTech Connect (OSTI)

    Mihalczo, John T [ORNL; Archer, Daniel E [ORNL; Wright, Michael C [ORNL; Mullens, James Allen [ORNL

    2007-09-01

    Measurements were performed with a single annular, stainless-steel-canned casting of uranium (93.17 wt% 235U) metal ( ~18 kg) to provide data to verify calculational methods for criticality safety. The measurements used a small portable DT generator with an embedded alpha detector to time and directionally tag the neutrons from the generator. The center of the time and directional tagged neutron beam was perpendicular to the axis of the casting. The radiation detectors were 1x1x6 in plastic scintillators encased in 0.635-cm-thick lead shields that were sensitive to neutrons above 1 MeV in energy. The detector lead shields were adjacent to the casting and the target spot of the generator was about 3.8 cm from the casting at the vertical center. The time distribution of the fission induced radiation was measured with respect to the source event by a fast (1GHz) processor. The measurements described in this paper also include time correlation measurements with a time tagged spontaneously fissioning 252Cf neutron source, both on the axis and on the surface of the casting. Measurements with both types of sources are compared. Measurements with the DT generator closely coupled with the HEU provide no more additional information than those with the Cf source closely coupled with the HEU and are complicated by the time and directionally tagged neutrons from the generator scattering between the walls and floor of the measurements room and the casting while still above detection thresholds.

  18. Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation

    SciTech Connect (OSTI)

    Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

    2008-07-01

    On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Re, Czech Republic, and discussed the return of their high-enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.

  19. Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water

    SciTech Connect (OSTI)

    Marshalkin, V. E. Povyshev, V. M.

    2015-12-15

    A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium–uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D{sub 2}O, H{sub 2}O) is proposed. The method is characterized by efficient breeding of the {sup 233}U isotope and safe reactor operation and is comparatively simple to implement.

  20. New generation enrichment monitoring technology for gas centrifuge enrichment plants

    SciTech Connect (OSTI)

    Ianakiev, Kiril D; Alexandrov, Boian S.; Boyer, Brian D.; Hill, Thomas R.; Macarthur, Duncan W.; Marks, Thomas; Moss, Calvin E.; Sheppard, Gregory A.; Swinhoe, Martyn T.

    2008-06-13

    The continuous enrichment monitor, developed and fielded in the 1990s by the International Atomic Energy Agency, provided a go-no-go capability to distinguish between UF{sub 6} containing low enriched (approximately 4% {sup 235}U) and highly enriched (above 20% {sup 235}U) uranium. This instrument used the 22-keV line from a {sup 109}Cd source as a transmission source to achieve a high sensitivity to the UF{sub 6} gas absorption. The 1.27-yr half-life required that the source be periodically replaced and the instrument recalibrated. The instrument's functionality and accuracy were limited by the fact that measured gas density and gas pressure were treated as confidential facility information. The modern safeguarding of a gas centrifuge enrichment plant producing low-enriched UF{sub 6} product aims toward a more quantitative flow and enrichment monitoring concept that sets new standards for accuracy stability, and confidence. An instrument must be accurate enough to detect the diversion of a significant quantity of material, have virtually zero false alarms, and protect the operator's proprietary process information. We discuss a new concept for advanced gas enrichment assay measurement technology. This design concept eliminates the need for the periodic replacement of a radioactive source as well as the need for maintenance by experts. Some initial experimental results will be presented.

  1. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    SciTech Connect (OSTI)

    Renfro, David G; Chandler, David; Cook, David Howard; Ilas, Germina; Jain, Prashant K; Valentine, Jennifer R

    2014-11-01

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present studies used current analytical tools to evaluate the various alternate designs for cycle length, scientific performance (e.g., neutron scattering), and steady-state and transient thermal performance using both safety limit and nominal parameter assumptions. The studies concluded that a new reference design combining a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone will allow successful conversion of HFIR. Future collaboration with the program will reveal whether the new reference design can be fabricated reliably and affordably. Following this feedback, additional studies using state-of-the-art developmental analytical tools are proposed to optimize the design of the fuel zone radial contour and the amount and location of both types of neutron absorbers to further flatten thermal peaks while maximizing the performance of the reactor.

  2. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    SciTech Connect (OSTI)

    Renfro, David; Chandler, David; Cook, David; Ilas, Germina; Jain, Prashant; Valentine, Jennifer

    2014-10-30

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present studies used current analytical tools to evaluate the various alternate designs for cycle length, scientific performance (e.g., neutron scattering), and steady-state and transient thermal performance using both safety limit and nominal parameter assumptions. The studies concluded that a new reference design combining a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone will allow successful conversion of HFIR. Future collaboration with the program will reveal whether the new reference design can be fabricated reliably and affordably. Following this feedback, additional studies using state-of-the-art developmental analytical tools are proposed to optimize the design of the fuel zone radial contour and the amount and location of both types of neutron absorbers to further flatten thermal peaks while maximizing the performance of the reactor.

  3. Enterprise Assessments Targeted Review of the Safety System Management of the Secondary Confinement System and Power Distribution Safety System at the Y-12 National Security Complex Highly Enriched Uranium Materials Facility – December 2015

    Broader source: Energy.gov [DOE]

    Targeted Review of the Safety System Management of the Secondary Confinement System and Power Distribution Safety System at the Y-12 National Security Complex Highly Enriched Uranium Materials Facility

  4. Experiments and Simulations of the Use of Time-Correlated Thermal Neutron Counting to Determine the Multiplication of an Assembly of Highly Enriched Uranium

    SciTech Connect (OSTI)

    David L. Chichester; Mathew T. Kinlaw; Scott M. Watson; Jeffrey M. Kalter; Eric C. Miller; William A. Noonan

    2014-11-01

    A series of experiments and numerical simulations using thermal-neutron time-correlated measurements has been performed to determine the neutron multiplication, M, of assemblies of highly enriched uranium available at Idaho National Laboratory. The experiments used up to 14.4 kg of highly-enriched uranium, including bare assemblies and assemblies reflected with high-density polyethylene, carbon steel, and tungsten. A small 252Cf source was used to initiate fission chains within the assembly. Both the experiments and the simulations used 6-channel and 8-channel detector systems, each consisting of 3He proportional counters moderated with polyethylene; data was recorded in list mode for analysis. 'True' multiplication values for each assembly were empirically derived using basic neutron production and loss values determined through simulation. A total of one-hundred and sixteen separate measurements were performed using fifty-seven unique measurement scenarios, the multiplication varied from 1.75 to 10.90. This paper presents the results of these comparisons and discusses differences among the various cases.

  5. Direct fissile assay of highly enriched UF/sub 6/ using random self-interrogation and neutron coincidence response

    SciTech Connect (OSTI)

    Stewart, J.E.; Menlove, H.O.

    1983-01-01

    A new nondestructive method for direct assay of /sup 235/U mass contained in Model 5A uranium hexafluoride (UF/sub 6/) product storage cylinders has been successfully tested in the laboratory and under field conditions. The technique employs passive neutron self-interrogation and uses the ratio of coincidences-to-totals counts as a measure of bulk fissile mass. The accuracy of the method is 6.8% (1 sigma) based on field measurements of 44 Model 5A cylinders, 11 of which were either only partially filled or contained reactor return material. The cylinders contained UF/sub 6/ with enrichments from 5.96% to 97.6%. Count times were 3 to 6 min depending on /sup 235/U mass. Samples ranged from below 1 kg to over 16 kg of /sup 235/U. Because the method relies primarily on fast neutron self-interrogation, complete sampling of the UF/sub 6/ takes place. This feature alleviates inhomogeneity problems and offers increased assurance of the presence of stated amounts of bulk fissile material as compared with current verification methods.

  6. DOE/NNSA Successfully Establishes Uranium Lease and Takeback...

    National Nuclear Security Administration (NNSA)

    the DOENNSA's ongoing support for the establishment of a domestic, commercial Mo-99 production capability that does not use proliferation-sensitive highly enriched uranium (HEU). ...

  7. Hight-Level Waste & Facilities Disposition

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Administration Highly Enriched Uranium Transparency Program November 13, 2013 The U.S. National Nuclear Security Administration's (NNSA) Highly Enriched Uranium (HEU) Transparency Program reduces nuclear risk by monitoring the conversion of 500 metric tons (MT) of Russian HEU, enough material for 20,000 nuclear weapons, into low enriched uranium (LEU). This LEU is put into peaceful use in the United States, generating nearly 10% of all U.S. electrical power. The HEU Purchase Agreement:

  8. Certification review of the DC-1 packaging for transport of HEU oxide

    SciTech Connect (OSTI)

    Primeau, S.J.; Thomas, G.R.

    1995-11-01

    The DC-1, a packaging for the transport of up to 19.7 kg of high-enriched uranium oxide, has been submitted for Department of Energy (DOE) Headquarters certification. The technical review for regulatory compliance is discussed from the reviewer`s standpoint, with emphasis on three notable aspects of the packaging design: (1) The packaging is the first submitted for DOE certification under the provisions of 10 CFR 71.55(c). (2) The containment vessels were fabricated in accordance with Section 8 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code; however, additional inspection requirements were imposed at the time of fabrication. Compliance with the requirements of Section 3 of the ASME Code is discussed. (3) A temporary adapter plate is installed between the lid and body of each containment vessel to facilitate periodic leak testing of the containment boundaries. This procedure is compared against the recommendations in the American National Standards Institute (ANSI) N14.5 Standard.

  9. Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

    SciTech Connect (OSTI)

    Kontogeorgakos, D.; Derstine, K.; Wright, A.; Bauer, T.; Stevens, J.

    2013-06-01

    The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO2 particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.

  10. GROTESQUE: Complex Geometric Arrangement of Unreflected HEU (93.15) Metal Pieces

    SciTech Connect (OSTI)

    Mackenzie L. Gorham; John D. Bess

    2011-09-01

    The GROTESQUE experiments were designed specifically to develop and test neutronics for the GEOM subroutine of the 05R code. Two complex arrangements of various highly enriched uranium metal cylinders, rectangular parallelepipeds, and spheres were arranged in a circular formation on a steel diaphragm. A centerpiece was raised remotely through a hole in the steel diaphragm to achieve criticality. The first arrangement consisted of five major units, each major unit consisting of a stack of smaller uranium pieces. The second arrangement utilized nine major units, again consisting of stacks of smaller uranium pieces. The 9-unit arrangement is the only experiment discussed in this evaluation, since the five stack experiment never achieved criticality. The 9-unit arrangement is shown in Figure 1.1. The experiments were performed at the Oak Ridge Critical Experiments Facility (ORCEF) in June 1964. The 9-unit configuration was later used as part of the development process for early versions of KENO and a model representing a variation of this experiment (Sample Problem 7: GROTESQUE without the Diaphragm) is released with modern versions of SCALE for testing the proper installation of the KENO module. An experimental report for the GROTESQUE experiment has not been published; however there are two publications that describe the experiment (References 1 and 2). A separate reportc discussing the conversion of the 05r model into a KENO model was published; however, the author did not consult with the experimenter for GROTESQUE. This report is considered unreliable (except for dimensions) by the experimenter and should not be used to obtain information pertinent to the GROTESQUE experiment. The Oak Ridge Critical Experiments Facility (ORCEF) Logbook 15r contains the primary documentation from the experimenter for this experiment. The GROTESQUE arrangement of nine major units was evaluated an determined to be an acceptable benchmark experiment.

  11. Overview of 2007 ANL progress for conversion of HEU- based Mo-99 production as part of the U.S. global threat reduction--conversion program

    SciTech Connect (OSTI)

    Vandegrift, George F.; Bakel, Allen J.; Thomas, Justin W.

    2008-07-15

    ANL effort is divided into five areas: (1) cooperation with Argentina to demonstrate the use of LEU-foil targets in alkaline-based processes, (2) cooperation with Indonesia in converting their HEU-based Cintichem process to LEU-foil targets, (3) technical assistance to two potential U.S. domestic suppliers (MURR and BWTX), (4) responding to the National Academies Study, and (5) participation in the IAEA CRP for Indigenous Mo-99 production. This paper presents highlights of these activities. A short description of how the dose emitted by spent HEU target material compared to spent fuel is also included. (author)

  12. Overview of 2007 ANL progress for conversion of HEU-based Mo-99 production as part of the U.S. Global Threat Reduction--Conversion program.

    SciTech Connect (OSTI)

    Vandegrift, G. F.; Bakel, A. J.; Thomas, J. W.

    2007-01-01

    ANL effort is divided into five areas: (1) cooperation with Argentina to demonstrate the use of LEU-foil targets in alkaline-based processes, (2) cooperation with Indonesia in converting their HEU-based Cintichem process to LEU-foil targets, (3) technical assistance to two potential U.S. domestic suppliers (MURR and BWTX), (4) responding to the National Academies Study, and (5) participation in the IAEA CRP for Indigenous Mo-99 production. This paper presents highlights of these activities. A short description of how the dose emitted by spent HEU target material compared to spent fuel is also included.

  13. Highly Enriched Uranium Materials Facility

    National Nuclear Security Administration (NNSA)

    Chuck Fleischmann (R-Tenn.) as well as NNSA Production Office Deputy Manager Teresa Robbins, Consolidated Nuclear Security President and Chief Executive Officer Jim Haynes, and...

  14. Enterprise Assessments Targeted Review of the Safety System Management of the Secondary Confinement System and Power Distribution Safety System at the Y-12 National Security Complex Highly Enriched Uranium Materials Facility … December 2015

    Energy Savers [EERE]

    Targeted Review of the Safety System Management of the Secondary Confinement System and Safety Significant Power Distribution System at the Y-12 National Security Complex Highly Enriched Uranium Materials Facility December 2015 Office of Nuclear Safety and Environmental Assessments Office of Environment, Safety and Health Assessments Office of Enterprise Assessments U.S. Department of Energy i Table of Contents Acronyms

  15. Belgium's Red Electrical Devils Win $1 Million for Innovative Inverter

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Security Administration Highly Enriched Uranium and Plutonium Removals March 24, 2014 Belgium has been a global leader in nonproliferation, working with the United States since 2006 to minimize highly enriched uranium (HEU) and plutonium inventories in Belgium through the return of a significant amount of HEU and plutonium to the United States. At the 2014 Nuclear Security Summit, the United States and Belgium announced the successful removal of all excess fresh HEU and plutonium from

  16. Safety aspects of gas centrifuge enrichment plants

    SciTech Connect (OSTI)

    Hansen, A.H.

    1987-01-01

    Uranium enrichment by gas centrifuge is a commercially proven, viable technology. Gas centrifuge enrichment plant operations pose hazards that are also found in other industries as well as unique hazards as a result of processing and handling uranium hexafluoride and the handling of enriched uranium. Hazards also found in other industries included those posed by the use of high-speed rotating equipment and equipment handling by use of heavy-duty cranes. Hazards from high-speed rotating equipment are associated with the operation of the gas centrifuges themselves and with the operation of the uranium hexafluoride compressors in the tail withdrawal system. These and related hazards are discussed. It is included that commercial gas centrifuge enrichment plants have been designed to operate safely.

  17. Stable Isotope Enrichment Capabilities at ORNL

    SciTech Connect (OSTI)

    Egle, Brian; Aaron, W Scott; Hart, Kevin J

    2013-01-01

    The Oak Ridge National Laboratory (ORNL) and the US Department of Energy Nuclear Physics Program have built a high-resolution Electromagnetic Isotope Separator (EMIS) as a prototype for reestablishing a US based enrichment capability for stable isotopes. ORNL has over 60 years of experience providing enriched stable isotopes and related technical services to the international accelerator target community, as well as medical, research, industrial, national security, and other communities. ORNL is investigating the combined use of electromagnetic and gas centrifuge isotope separation technologies to provide research quantities (milligram to several kilograms) of enriched stable isotopes. In preparation for implementing a larger scale production facility, a 10 mA high-resolution EMIS prototype has been built and tested. Initial testing of the device has simultaneously collected greater than 98% enriched samples of all the molybdenum isotopes from natural abundance feedstock.

  18. Possibility of nuclear pumped laser experiment using low enriched uranium

    SciTech Connect (OSTI)

    Obara, Toru; Takezawa, Hiroki [Center for Research into Innovative Nuclear Energy Systems Tokyo Institute of Technology 2-12-1-N1-19, Ookayama Meguro-ku, Tokyo 152-8550 (Japan)

    2012-06-06

    Possibility to perform experiments for nuclear pumped laser oscillation by using low enriched uranium is investigated. Kinetic analyses are performed for two types of reactor design, one is using highly enriched uranium and the other is using low enriched uranium. The reactor design is based on the experiment reactor in IPPE. The results show the oscillation of nuclear pumped laser in the case of low enriched uranium reactor is also possible. The use of low enriched uranium in the experiment will make experiment easier.

  19. Advanced Multiphysics Thermal-Hydraulics Models for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Jain, Prashant K; Freels, James D

    2015-01-01

    Engineering design studies to determine the feasibility of converting the High Flux Isotope Reactor (HFIR) from using highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL). This work is part of an effort sponsored by the US Department of Energy (DOE) Reactor Conversion Program. HFIR is a very high flux pressurized light-water-cooled and moderated flux-trap type research reactor. HFIR s current missions are to support neutron scattering experiments, isotope production, and materials irradiation, including neutron activation analysis. Advanced three-dimensional multiphysics models of HFIR fuel were developed in COMSOL software for safety basis (worst case) operating conditions. Several types of physics including multilayer heat conduction, conjugate heat transfer, turbulent flows (RANS model) and structural mechanics were combined and solved for HFIR s inner and outer fuel elements. Alternate design features of the new LEU fuel were evaluated using these multiphysics models. This work led to a new, preliminary reference LEU design that combines a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone. Preliminary results of estimated thermal safety margins are presented. Fuel design studies and model enhancement continue.

  20. The Enriched Xenon Observatory

    SciTech Connect (OSTI)

    Dolinski, M. J. [Stanford University Physics Department, 382 Via Pueblo Mall, Stanford, CA 94305-4060 (United States)

    2009-12-17

    The Enriched Xenon Observatory (EXO) experiment will search for neutrinoless double beta decay of {sup 136}Xe. The EXO Collaboration is actively pursuing both liquid-phase and gas-phase Xe detector technologies with scalability to the ton-scale. The search for neutrinoless double beta decay of {sup 136}Xe is especially attractive because of the possibility of tagging the resulting Ba daughter ion, eliminating all sources of background other than the two neutrino decay mode. EXO-200, the first phase of the project, is a liquid Xe time projection chamber with 200 kg of Xe enriched to 80% in {sup 136}Xe. EXO-200, which does not include Ba-tagging, will begin taking data in 2009, with two-year sensitivity to the half-life for neutrinoless double beta decay of 6.4x10{sup 25} years. This corresponds to an effective Majorana neutrino mass of 0.13 to 0.19 eV.

  1. Conversion and enrichment in the Soviet Union

    SciTech Connect (OSTI)

    1991-04-01

    In the Soviet Union, just as in the West, the civilian nuclear industry emerged from research work undertaken for nuclear weapons development. At first, researchers tried various techniques for physical separation of uranium isotopes: electromagnetic and molecular-kinetic thermo-diffusion methods; gaseous diffusion; and centrifuge methods. All of those methods, which are based primarily on differences in the atomic mass of uranium isotopes, called for extensive research and the development of new, technically unprecedented equipment. Gradually gaseous diffusion and gas centrifuge technology became recognized as most feasible for industrial use, so research on other methods was terminated. Industrial-scale uranium enrichment in the Soviet Union began in 1949 using the gaseous diffusion method; by the early 1960s, centrifuge technology was in use on an industrial scale. All Soviet production of highly-enriched, weapons-grade uranium was halted in 1987. The Soviet Union now has four enrichment plants in operation (at classified locations), solely for civilian nuclear power needs. All four enrichment plants have centrifuge modules, and enrichment provided by gaseous diffusion accounts for less than 5% of their total output. Two of the four enrichment plants also incorporate facilities for conversion to uranium hexafluoride (UF{sub 6}).

  2. Overview of enrichment plant safeguards

    SciTech Connect (OSTI)

    Swindle, D.W. Jr.; Wheeler, L.E.

    1982-01-01

    The relationship of enrichment plant safeguards to US nonproliferation objectives and to the operation and management of enrichment facilities is reviewed. During the review, the major components of both domestic and international safeguards systems for enrichment plants are discussed. In discussing domestic safeguards systems, examples of the technology currently in use to support nuclear materials accountability are described including the measurement methods, procedures and equipment used for weighing, sampling, chemical and isotopic analyses and nondestructive assay techniques. Also discussed is how the information obtained as part of the nuclear material accountancy task is useful to enrichment plant operations. International material accountancy verification and containment/surveillance concepts for enrichment plants are discussed, and the technologies presently being developed for international safeguards in enrichment plants are identified and the current development status is reported.

  3. Assuaging Nuclear Energy Risks: The Angarsk International Uranium Enrichment Center

    SciTech Connect (OSTI)

    Myers, Astasia

    2011-06-28

    The recent nuclear renaissance has motivated many countries, especially developing nations, to plan and build nuclear power reactors. However, domestic low enriched uranium demands may trigger nations to construct indigenous enrichment facilities, which could be redirected to fabricate high enriched uranium for nuclear weapons. The potential advantages of establishing multinational uranium enrichment sites are numerous including increased low enrichment uranium access with decreased nuclear proliferation risks. While multinational nuclear initiatives have been discussed, Russia is the first nation to actualize this concept with their Angarsk International Uranium Enrichment Center (IUEC). This paper provides an overview of the historical and modern context of the multinational nuclear fuel cycle as well as the evolution of Russia's IUEC, which exemplifies how international fuel cycle cooperation is an alternative to domestic facilities.

  4. Phosphorus Enrichment as a New Composition in the Solid Electrolyte Interphase of High-Voltage Cathodes and Its Effects on Battery Cycling

    SciTech Connect (OSTI)

    Yan, Pengfei; Zheng, Jianming; Kuppan, Saravanan; Li, Qiuyan; Lv, Dongping; Xiao, Jie; Chen, Guoying; Zhang, Jiguang; Wang, Chong M.

    2015-11-10

    Immersion of a solid into liquid often leads to the modification of both the structure and chemistry of surface of the solid, which subsequently affects the chemical and physical properties of the system. For the case of the rechargeable lithium ion battery, such a surface modification is termed as solid electrolyte interphase (SEI) layer, which has been perceived to play critical role for the stable operation of the batteries. However, the structure and chemical composition of SEI layer and its spatial distribution and dependence on the battery operating condition remain unclear. By using aberration corrected scanning transmission electron microscopy coupled with ultra-high sensitive energy dispersive x-ray spectroscopy, we probed the structure and chemistry of SEI layer on several high voltage cathodes. We show that layer-structured cathodes, when cycled at a high cut off voltage, can form a P-rich SEI layer on their surface, which is a direct evidence of Li-salt (LiPF6) decomposition. Our systematical investigations indicate such cathode/Li-salt side reaction shows strong dependence on structure of the cathode materials, operating voltage and temperature, indicating the feasibility of SEI engineering. These findings provide us valuable insights into the complex interface between the high-voltage cathode and the electrolyte.

  5. Uranium Enrichment Decontamination and Decommissioning Fund's...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2008 and 2007 Financial Statement Audit, OAS-FS-10-05 Uranium Enrichment Decontamination and...

  6. Mo-99

    National Nuclear Security Administration (NNSA)

    its project for domestic production of molybdenum-99 (Mo-99) without highly enriched uranium (HEU).

    Mo-99 is the parent isotope of technetium-99m, which is the most widely...

  7. EIS-0279: Amended Record of Decision | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    target materials from Canada that contain liquid Highly Enriched Uranium (HEU) of U.S. origin). PDF icon EIS-0279-AmenROD-2013.pdf More Documents & Publications EIS-0279: Record of ...

  8. An Optically Stimulated Luminescence Uranium Enrichment Monitor

    SciTech Connect (OSTI)

    Miller, Steven D.; Tanner, Jennifer E.; Simmons, Kevin L.; Conrady, Matthew M.; Benz, Jacob M.; Greenfield, Bryce A.

    2010-08-11

    The Pacific Northwest National Laboratory (PNNL) has pioneered the use of Optically Stimulated Luminescence (OSL) technology for use in personnel dosimetry and high dose radiation processing dosimetry. PNNL has developed and patented an alumina-based OSL dosimeter that is being used by the majority of medical X-ray and imaging technicians worldwide. PNNL has conceived of using OSL technology to passively measure the level of UF6 enrichment by attaching the prototype OSL monitor to pipes containing UF6 gas within an enrichment facility. The prototype OSL UF6 monitor utilizes a two-element approach with the first element open and unfiltered to measure both the low energy and high energy gammas from the UF6, while the second element utilizes a 3-mm thick tungsten filter to eliminate the low energy gammas and pass only the high energy gammas from the UF6. By placing a control monitor in the room away from the UF6 pipes and other ionizing radiation sources, the control readings can be subtracted from the UF6 pipe monitor measurements. The ratio of the shielded to the unshielded net measurements provides a means to estimate the level of uranium enrichment. PNNL has replaced the commercially available MicroStar alumina-based dosimeter elements with a composite of polyethylene plastic, high-Z glass powder, and BaFBr:Eu OSL phosphor powder at various concentrations. The high-Z glass was added in an attempt to raise the average Z of the composite dosimeter and increase the response. Additionally, since BaFBr:Eu OSL phosphor is optimally excited and emits light at different wavelengths compared to alumina, the commercially available MicroStar reader was modified for reading BaFBr:Eu in a parallel effort to increase reader sensitivity. PNNL will present the design and performance of our novel OSL uranium enrichment monitor based on a combination of laboratory and UF6 test loop measurements. PNNL will also report on the optimization effort to achieve the highest possible performance from both the OSL enrichment monitor and the new custom OSL reader modified for this application. This project has been supported by the US Department of Energys National Nuclear Security Administrations Office of Dismantlement and Transparency (DOE/NNSA/NA-241).

  9. highly enriched uranium | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    This Site Budget IG Web Policy Privacy No Fear Act Accessibility FOIA Sitemap Federal Government The White House DOE.gov USA.gov Jobs Apply for Our Jobs Our Jobs Working at NNSA...

  10. Highly Enriched Uranium Transparency Program | National Nuclear...

    National Nuclear Security Administration (NNSA)

    This Site Budget IG Web Policy Privacy No Fear Act Accessibility FOIA Sitemap Federal Government The White House DOE.gov USA.gov Jobs Apply for Our Jobs Our Jobs Working at NNSA...

  11. Highly Enriched Uranium Materials Facility | National Nuclear...

    National Nuclear Security Administration (NNSA)

    This Site Budget IG Web Policy Privacy No Fear Act Accessibility FOIA Sitemap Federal Government The White House DOE.gov USA.gov Jobs Apply for Our Jobs Our Jobs Working at NNSA...

  12. Ulrich Wiesner > Spencer T. Olin Professor of Engineering

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Nuclear Security Administration US, Kazakhstan Cooperate to Eliminate Highly Enriched Uranium January 07, 2015 WASHINGTON D.C - The Department of Energy's National Nuclear Security Administration (DOE/NNSA) announced today the removal of 36 kilograms (approximately 80 pounds) of highly enriched uranium (HEU) spent fuel from the Institute of Nuclear Physics (INP) in Almaty, Kazakhstan. The HEU was transported via two air shipments to a secure facility in Russia for permanent disposition. This

  13. russia | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    russia NNSA Partnership Successfully Removes All Remaining HEU from Uzbekistan WASHINGTON, DC - Today, the Department of Energy's National Nuclear Security Administration (DOE/NNSA) announced the successful return of the final 5 kilograms (approximately 11 pounds) of highly enriched uranium (HEU) spent fuel from the IIN-3M "Foton" research reactor in Tashkent, Uzbekistan to... US, Kazakhstan Cooperate to Eliminate Highly Enriched Uranium WASHINGTON D.C - The Department of Energy's

  14. request.pdf

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    removal US, Kazakhstan Cooperate to Eliminate Highly Enriched Uranium WASHINGTON D.C - The Department of Energy's National Nuclear Security Administration (DOE/NNSA) announced today the removal of 36 kilograms (approximately 80 pounds) of highly enriched uranium (HEU) spent fuel from the Institute of Nuclear Physics (INP) in Almaty, Kazakhstan. The HEU was

    render safe Nuclear Forensics The National Technical Nuclear Forensics (NTNF) program is a Homeland Security Council and National Security

  15. Iyer, Woldegabriel recognized for helping New Mexico small businesses

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Security Administration Highly Enriched Uranium and Plutonium Removals March 24, 2014 Italy has been a global leader in nuclear nonproliferation, working with the United States since 1997 to eliminate more than 100 kilograms of highly enriched uranium (HEU) and separated plutonium. At the 2014 Nuclear Security Summit, the United States and Italy announced the successful removal of all eligible fresh HEU and plutonium from Italy. These shipments were completed via a joint effort between the

  16. Enrichment of light hydrocarbon mixture

    DOE Patents [OSTI]

    Yang, Dali; Devlin, David; Barbero, Robert S.; Carrera, Martin E.; Colling, Craig W.

    2011-11-29

    Light hydrocarbon enrichment is accomplished using a vertically oriented distillation column having a plurality of vertically oriented, nonselective micro/mesoporous hollow fibers. Vapor having, for example, both propylene and propane is sent upward through the distillation column in between the hollow fibers. Vapor exits neat the top of the column and is condensed to form a liquid phase that is directed back downward through the lumen of the hollow fibers. As vapor continues to ascend and liquid continues to countercurrently descend, the liquid at the bottom of the column becomes enriched in a higher boiling point, light hydrocarbon (propane, for example) and the vapor at the top becomes enriched in a lower boiling point light hydrocarbon (propylene, for example). The hollow fiber becomes wetted with liquid during the process.

  17. Enrichment of light hydrocarbon mixture

    DOE Patents [OSTI]

    Yang; Dali; Devlin, David; Barbero, Robert S.; Carrera, Martin E.; Colling, Craig W.

    2010-08-10

    Light hydrocarbon enrichment is accomplished using a vertically oriented distillation column having a plurality of vertically oriented, nonselective micro/mesoporous hollow fibers. Vapor having, for example, both propylene and propane is sent upward through the distillation column in between the hollow fibers. Vapor exits neat the top of the column and is condensed to form a liquid phase that is directed back downward through the lumen of the hollow fibers. As vapor continues to ascend and liquid continues to countercurrently descend, the liquid at the bottom of the column becomes enriched in a higher boiling point, light hydrocarbon (propane, for example) and the vapor at the top becomes enriched in a lower boiling point light hydrocarbon (propylene, for example). The hollow fiber becomes wetted with liquid during the process.

  18. AVLIS enrichment of medical isotopes

    SciTech Connect (OSTI)

    Haynam, C.A.; Scheibner, K.F.; Stern, R.C.; Worden, E.F.

    1996-12-31

    Under the Sponsorship of the United states Enrichment Corporation (USEC), we are currently investigating the large scale separation of several isotopes of medical interest using atomic vapor isotope separation (AVLIS). This work includes analysis and experiments in the enrichment of thallium 203 as a precursor to the production of thallium 201 used in cardiac imaging following heart attacks, on the stripping of strontium 84 from natural strontium as precursor to the production of strontium 89, and on the stripping of lead 210 from lead used in integrated circuits to reduce the number of alpha particle induced logic errors.

  19. Jefferson Lab welcomes students, teachers for summer internship, enrichment

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    program | Jefferson Lab welcomes students, teachers for summer internship, enrichment program 2003 Education Poster Session Jefferson Lab welcomes students, teachers for summer internship, enrichment program July 28, 2004 Newport News, VA. - As schools close for the summer, the number of teachers and high school and college students at the Department of Energy's Jefferson Lab in Newport News, Va., multiplies. They come to participate in a variety of innovative, educational, science-based

  20. SAGE Application Process

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Spring 2014 National Nuclear Security Administration ENERGY U.S. DEPARTMENT OF NIS Observes Final Shipment of LEU from Russian Weapons HEU BY GREG DWYER A s NIS experts watched, the final 40 cylinders of low enriched uranium (LEU) derived from Russian highly enriched uranium (HEU) crossed the rail of the Atlantic Navigator at the Port of Saint Petersburg, Russia on December 14, 2013. These cylinders were the last of nearly 10,000 LEU cylinders delivered over the past Elementary! A Nuclear

  1. SAFEGUARD AND SECURE CONTROL VERIFY POLICY

    National Nuclear Security Administration (NNSA)

    Spring 2014 National Nuclear Security Administration ENERGY U.S. DEPARTMENT OF NIS Observes Final Shipment of LEU from Russian Weapons HEU BY GREG DWYER A s NIS experts watched, the final 40 cylinders of low enriched uranium (LEU) derived from Russian highly enriched uranium (HEU) crossed the rail of the Atlantic Navigator at the Port of Saint Petersburg, Russia on December 14, 2013. These cylinders were the last of nearly 10,000 LEU cylinders delivered over the past Elementary! A Nuclear

  2. Argonne supports Poland's NCNR in converting reactor to LEU | Argonne

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    National Laboratory supports Poland's NCNR in converting reactor to LEU October 23, 2012 Tweet EmailPrint With assistance from the U.S. Department of Energy's National Nuclear Security Administration (NNSA), the MARIA civilian research reactor in Poland was converted from operation with highly enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel on September 24, 2012. The irradiated HEU fuel removed from MARIA will be returned to Russia. NNSA assisted this reactor conversion

  3. A Robust Infrastructure Design for Gas Centrifuge Enrichment Plant Unattended Online Enrichment Monitoring

    SciTech Connect (OSTI)

    Younkin, James R; Rowe, Nathan C; Garner, James R

    2012-01-01

    An online enrichment monitor (OLEM) is being developed to continuously measure the relative isotopic composition of UF6 in the unit header pipes of a gas centrifuge enrichment plant (GCEP). From a safeguards perspective, OLEM will provide early detection of a facility being misused for production of highly enriched uranium. OLEM may also reduce the number of samples collected for destructive assay and if coupled with load cell monitoring can provide isotope mass balance verification. The OLEM design includes power and network connections for continuous monitoring of the UF6 enrichment and state of health of the instrument. Monitoring the enrichment on all header pipes at a typical GCEP could require OLEM detectors on each of the product, tails, and feed header pipes. If there are eight process units, up to 24 detectors may be required at a modern GCEP. Distant locations, harsh industrial environments, and safeguards continuity of knowledge requirements all place certain demands on the network robustness and power reliability. This paper describes the infrastructure and architecture of an OLEM system based on OLEM collection nodes on the unit header pipes and power and network support nodes for groupings of the collection nodes. A redundant, self-healing communications network, distributed backup power, and a secure communications methodology. Two candidate technologies being considered for secure communications are the Object Linking and Embedding for Process Control Unified Architecture cross-platform, service-oriented architecture model for process control communications and the emerging IAEA Real-time And INtegrated STream-Oriented Remote Monitoring (RAINSTORM) framework to provide the common secure communication infrastructure for remote, unattended monitoring systems. The proposed infrastructure design offers modular, commercial components, plug-and-play extensibility for GCEP deployments, and is intended to meet the guidelines and requirements for unattended and remotely monitored safeguards systems.

  4. Prompt Neutron Lifetime for the NBSR Reactor

    SciTech Connect (OSTI)

    Hanson, A.L.; Diamond, D.

    2012-06-24

    In preparation for the proposed conversion of the National Institute of Standards and Technology (NIST) research reactor (NBSR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, certain point kinetics parameters must be calculated. We report here values of the prompt neutron lifetime that have been calculated using three independent methods. All three sets of calculations demonstrate that the prompt neutron lifetime is shorter for the LEU fuel when compared to the HEU fuel and longer for the equilibrium end-of-cycle (EOC) condition when compared to the equilibrium startup (SU) condition for both the HEU and LEU fuels.

  5. Peak fitting applied to low-resolution enrichment measurements

    SciTech Connect (OSTI)

    Bracken, D.; McKown, T.; Sprinkle, J.K. Jr.; Gunnink, R.; Kartoshov, M.; Kuropatwinski, J.; Raphina, G.; Sokolov, G.

    1998-12-01

    Materials accounting at bulk processing facilities that handle low enriched uranium consists primarily of weight and uranium enrichment measurements. Most low enriched uranium processing facilities draw separate materials balances for each enrichment handled at the facility. The enrichment measurement determines the isotopic abundance of the {sup 235}U, thereby determining the proper strata for the item, while the weight measurement generates the primary accounting value for the item. Enrichment measurements using the passive gamma radiation from uranium were developed for use in US facilities a few decades ago. In the US, the use of low-resolution detectors was favored because they cost less, are lighter and more robust, and don`t require the use of liquid nitrogen. When these techniques were exported to Europe, however, difficulties were encountered. Two of the possible root causes were discovered to be inaccurate knowledge of the container wall thickness and higher levels of minor isotopes of uranium introduced by the use of reactor returns in the enrichment plants. the minor isotopes cause an increase in the Compton continuum under the 185.7 keV assay peak and the observance of interfering 238.6 keV gamma rays. The solution selected to address these problems was to rely on the slower, more costly, high-resolution gamma ray detectors when the low-resolution method failed. Recently, these gamma ray based enrichment measurement techniques have been applied to Russian origin material. The presence of interfering gamma radiation from minor isotopes was confirmed. However, with the advent of fast portable computers, it is now possible to apply more sophisticated analysis techniques to the low-resolution data in the field. Explicit corrections for Compton background, gamma rays from {sup 236}U daughters, and the attenuation caused by thick containers can be part of the least squares fitting routine. Preliminary results from field measurements in Kazakhstan will be discussed.

  6. Health and safety considerations for U. S. monitors in the Russian transparency program.

    SciTech Connect (OSTI)

    Boggs, C. J.

    1998-10-22

    In 1993 the US and the Russian Federation signed an agreement allowing the US to purchase highly enriched uranium (HEU) from Russia over a 20-year period. This Highly Enriched Uranium Purchase Agreement permits the purchase of 500 metric tons of HEU from dismantled Russian nuclear weapons in the form of low-enriched uranium (LEU) for use as power reactor fuel in the US. Under the HEU Agreement, the US and Russia are cooperating in a ''Transparency Program'' to ensure that arms control and nonproliferation objectives are being met. The Transparency Program measures, which are a departure from traditional, intrusive measures of verification, include sending individuals from the US to Russia to monitor the processing of the HEU.

  7. Enrichment Assay Methods for a UF6 Cylinder Verification Station

    SciTech Connect (OSTI)

    Smith, Leon E.; Jordan, David V.; Misner, Alex C.; Mace, Emily K.; Orton, Christopher R.

    2010-11-30

    International Atomic Energy Agency (IAEA) inspectors currently perform periodic inspections at uranium enrichment plants to verify UF6 cylinder enrichment declarations. Measurements are typically performed with handheld high-resolution sensors on a sampling of cylinders taken to be representative of the facility’s entire cylinder inventory. These enrichment assay methods interrogate only a small fraction of the total cylinder volume, and are time-consuming and expensive to execute for inspectors. Pacific Northwest National Laboratory (PNNL) is developing an unattended measurement system capable of automated enrichment measurements over the full volume of Type 30B and Type 48 cylinders. This Integrated Cylinder Verification System (ICVS) could be located at key measurement points to positively identify each cylinder, measure its mass and enrichment, store the collected data in a secure database, and maintain continuity of knowledge on measured cylinders until IAEA inspector arrival. The focus of this paper is the development of nondestructive assay (NDA) methods that combine “traditional” enrichment signatures (e.g. 185-keV emission from U-235) and more-penetrating “non-traditional” signatures (e.g. high-energy neutron-induced gamma rays spawned primarily from U-234 alpha emission) collected by medium-resolution gamma-ray spectrometers (i.e. sodium iodide or lanthanum bromide). The potential of these NDA methods for the automated assay of feed, tail and product cylinders is explored through MCNP modeling and with field measurements on a cylinder population ranging from 0.2% to 5% in U-235 enrichment.

  8. Novel Membranes and Processes for Oxygen Enrichment

    SciTech Connect (OSTI)

    Lin, Haiqing

    2011-11-15

    The overall goal of this project is to develop a membrane process that produces air containing 25-35% oxygen, at a cost of $25-40/ton of equivalent pure oxygen (EPO2). Oxygen-enriched air at such a low cost will allow existing air-fueled furnaces to be converted economically to oxygen-enriched furnaces, which in turn will improve the economic and energy efficiency of combustion processes significantly, and reduce the cost of CO{sub 2} capture and sequestration from flue gases throughout the U.S. manufacturing industries. During the 12-month Concept Definition project: We identified a series of perfluoropolymers (PFPs) with promising oxygen/nitrogen separation properties, which were successfully made into thin film composite membranes. The membranes showed oxygen permeance as high as 1,200 gpu and oxygen/nitrogen selectivity of 3.0, and the permeance and selectivity were stable over the time period tested (60 days). We successfully scaled up the production of high-flux PFP-based membranes, using MTR's commercial coaters. Two bench-scale spiral-wound modules with countercurrent designs were made and parametric tests were performed to understand the effect of feed flow rate and pressure, permeate pressure and sweep flow rate on the membrane module separation properties. At various operating conditions that modeled potential industrial operating conditions, the module separation properties were similar to the pure-gas separation properties in the membrane stamps. We also identified and synthesized new polymers [including polymers of intrinsic microporosity (PIMs) and polyimides] with higher oxygen/nitrogen selectivity (3.5-5.0) than the PFPs, and made these polymers into thin film composite membranes. However, these membranes were susceptible to severe aging; pure-gas permeance decreased nearly six-fold within two weeks, making them impractical for industrial applications of oxygen enrichment. We tested the effect of oxygen-enriched air on NO{sub x} emissions using a Bloom baffle burner at GTI. The results are positive and confirm that oxygen-enriched combustion can be carried out without producing higher levels of NOx than normal air firing, if lancing of combustion air is used and the excess air levels are controlled. A simple economic study shows that the membrane processes can produce O{sub 2} at less than $40/ton EPO{sub 2} and an energy cost of 1.1-1.5 MMBtu/ton EPO{sub 2}, which are very favorable compared with conventional technologies such as cryogenics and vacuum pressure swing adsorption processes. The benefits of integrated membrane processes/combustion process trains have been evaluated, and show good savings in process costs and energy consumption, as well as reduced CO{sub 2} emissions. For example, if air containing 30% oxygen is used in natural gas furnaces, the net natural gas savings are an estimated 18% at a burner temperature of 2,500 F, and 32% at a burner temperature of 3,000 F. With a 20% market penetration of membrane-based oxygen-enriched combustion in all combustion processes by 2020, the energy savings would be 414-736 TBtu/y in the U.S. The comparable net cost savings are estimated at $1.2-2.1 billion per year by 2020, calculated as the value of fuel savings subtracted from the cost of oxygen production. The fuel savings of 18%-32% by the membrane/oxygen-enriched combustion corresponds to an 18%-32% reduction in CO{sub 2} emissions, or 23-40 MM ton/y less CO{sub 2} from natural gas-fired furnaces by 2020. In summary, results from this project (Concept Definition phase) are highly promising and clearly demonstrate that membrane processes can produce oxygen-enriched air in a low cost manner that will lower operating costs and energy consumption in industrial combustion processes. Future work will focus on proof-of-concept bench-scale demonstration in the laboratory.

  9. Automated UF6 Cylinder Enrichment Assay: Status of the Hybrid Enrichment Verification Array (HEVA) Project: POTAS Phase II

    SciTech Connect (OSTI)

    Jordan, David V.; Orton, Christopher R.; Mace, Emily K.; McDonald, Benjamin S.; Kulisek, Jonathan A.; Smith, Leon E.

    2012-06-01

    Pacific Northwest National Laboratory (PNNL) intends to automate the UF6 cylinder nondestructive assay (NDA) verification currently performed by the International Atomic Energy Agency (IAEA) at enrichment plants. PNNL is proposing the installation of a portal monitor at a key measurement point to positively identify each cylinder, measure its mass and enrichment, store the data along with operator inputs in a secure database, and maintain continuity of knowledge on measured cylinders until inspector arrival. This report summarizes the status of the research and development of an enrichment assay methodology supporting the cylinder verification concept. The enrichment assay approach exploits a hybrid of two passively-detected ionizing-radiation signatures: the traditional enrichment meter signature (186-keV photon peak area) and a non-traditional signature, manifested in the high-energy (3 to 8 MeV) gamma-ray continuum, generated by neutron emission from UF6. PNNL has designed, fabricated, and field-tested several prototype assay sensor packages in an effort to demonstrate proof-of-principle for the hybrid assay approach, quantify the expected assay precision for various categories of cylinder contents, and assess the potential for unsupervised deployment of the technology in a portal-monitor form factor. We refer to recent sensor-package prototypes as the Hybrid Enrichment Verification Array (HEVA). The report provides an overview of the assay signatures and summarizes the results of several HEVA field measurement campaigns on populations of Type 30B UF6 cylinders containing low-enriched uranium (LEU), natural uranium (NU), and depleted uranium (DU). Approaches to performance optimization of the assay technique via radiation transport modeling are briefly described, as are spectroscopic and data-analysis algorithms.

  10. Categorical Exclusion 4577: Lithium Isotope Separation & Enrichment...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Lithium Isotope Separation & Enrichment Technologies (4577) Program or Field Office: Y-12 Site Office Location(s) (CityCountyState): Oak Ridge, Anderson County, Tennessee...

  11. Use of MCNP + GADRAS in Generating More Realistic Gamma-Ray Spectra for Plutonium and HEU Objects

    SciTech Connect (OSTI)

    Rawool-Sullivan, Mohini; Mattingly, John; Mitchell, Dean

    2012-08-07

    The ability to accurately simulate high-resolution gamma spectra from materials that emit both neutrons and gammas is very important to the analysis of special nuclear materials (SNM), e.g., uranium and plutonium. One approach under consideration has been to combine MCNP and GADRAS. This approach is expected to generate more accurate gamma ray spectra for complex three-dimensional geometries than can be obtained from one-dimensional deterministic transport simulations (e.g., ONEDANT). This presentation describes application of combining MCNP and GADRAS in simulating plutonium and uranium spectra.

  12. Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

    SciTech Connect (OSTI)

    Gauntt, Randall O.; Ross, Kyle W.; Smith, James Dean; Longmire, Pamela

    2010-04-01

    The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.

  13. World enrichment services market 1990-2005

    SciTech Connect (OSTI)

    1990-08-01

    Growth in world enrichment capacity, already in oversupply, will lead to a very competitive enrichment services market by the second half of the 1990s. Three of the four primary enrichment suppliers (USDOE, Eurodif, and Urenco) already have the capacity to produce 33 million SWU per year. Explorts from the Soviet Union and the People`s Republic of China (PRC) currently make available an additional six million SWU per year, and that figure could rise substantially. With additional supply capability expected from China, the Soviet Union, Louisiana Energy Services (LES) and Isotope Technologies (ITI), and the increased capacity of Urenco, and possibly even AVLIS from DOE, enrichment supply capability could exceed 46 million SWU per year by the year 2000. Yet annual enrichment requirements are only estimated to grow from 23.5 million SWU in 1990, to 28.9 million SWU by 2000. Total unfilled enrichment requirements will rise significantly in the second half of the 1990s, particularly from US utilities, creating sales opportunities for which suppliers will compete aggressively. These factors foretell a very competitive market in which sellers will offer low prices and flexible contracts. The anticipation of such strong competition also raises the question of which enrichment technology will succeed, and puts tremendous pressure on all suppliers to find cost-effective means of production as quickly as possible.

  14. Uranium enrichment: heading for a cliff

    SciTech Connect (OSTI)

    Norman, C.

    1987-05-22

    Thanks to drastic cost cutting in the past 2 years, US enrichment plants now have the lowest cost production in the world, but US prices are still higher than those of overseas competitors because the business is paying for past mistakes. The most serious difficulty is that the Department of Energy (DOE), which owns and operates the US enrichment enterprise, is paying more than $500 million a year to the Tennessee Valley Authority (TVA) for electricity it once thought it would need but no longer requires. Another is that billions of dollars were spent in the 1970s and early 1980s to build new capacity that is now not needed. As a result, the enrichment enterprise has accumulated a debt to the US Treasury that the General Accounting Office (GAO) estimates at $8.8 billion. This paper presents the background and current debate in Congress about the difficulties facing the enrichment industry. In the midst of this debate over the future of the enterprise, the development of the next generation of enrichment technology is being placed in jeopardy. Known as atomic vapor laser isotope separation, or AVLIS, the process was viewed as the key to the long-term competitiveness of US enrichment. As the federal deficit mounted, however, funding for the AVLIS program was cut back and the timetable was stretched out. The US enrichment program has reached the point at which Congress will be forced to make some politically difficult decisions.

  15. Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Agreement | Department of Energy Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance Agreement Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance Agreement Toxic Substances Control Act (TSCA) Uranium Enrichment Federal Facility Compliance Agreement establishes a plan to bring DOE's Uranium Enrichment Plants (and support facilities) located in Portsmouth, Ohio and Paducah, Kentucky and DOE's former Uranium Enrichment Plant (and support

  16. The IMCA: A field instrument for uranium enrichment measurements

    SciTech Connect (OSTI)

    Gardner, G.H.; Koskelo, M.; Moeslinger, M.; Mayer, R.L. II; McGinnis, B.R.; Wishard, B.

    1996-12-31

    The IMCA (Inspection Multi-Channel Analyzer) is a portable gamma-ray spectrometer designed to measure the enrichment of uranium either in a laboratory or in the field. The IMCA consists of a Canberra InSpector Multi-Channel Analyzer, sodium iodide or a planar germanium detector, and special application software. The system possesses a high degree of automation. The IMCA uses the uranium enrichment meter principle, and is designed to meet the International Atomic Energy Agency (IAEA) requirements for the verification of enriched uranium materials. The IMCA is available with MGA plutonium isotopic analysis software or MGAU uranium analysis software as well. In this paper, the authors present a detailed description of the hardware and software of the IMCA system, as well as results from preliminary measurements testing compliance of IMCA with IAEA requirements using uranium standards and UF6 cylinders. Measurements performed on UF6 cylinders in the field under variable environmental conditions (temperatures ranging from 0 to 35 C) have shown that good results can be achieved. The enrichment of UF6 contained in the cylinder is determined by using calibration constants generated from an instrument calibration, using traceable uranium oxide standards, performed in the laboratory under controlled environmental conditions. The IMCA software is designed to make the necessary matrix and container corrections to ensure that accurate results are achieved in the field.

  17. Enrichment Assay Methods Development for the Integrated Cylinder Verification System

    SciTech Connect (OSTI)

    Smith, Leon E.; Misner, Alex C.; Hatchell, Brian K.; Curtis, Michael M.

    2009-10-22

    International Atomic Energy Agency (IAEA) inspectors currently perform periodic inspections at uranium enrichment plants to verify UF6 cylinder enrichment declarations. Measurements are typically performed with handheld high-resolution sensors on a sampling of cylinders taken to be representative of the facility's entire product-cylinder inventory. Pacific Northwest National Laboratory (PNNL) is developing a concept to automate the verification of enrichment plant cylinders to enable 100 percent product-cylinder verification and potentially, mass-balance calculations on the facility as a whole (by also measuring feed and tails cylinders). The Integrated Cylinder Verification System (ICVS) could be located at key measurement points to positively identify each cylinder, measure its mass and enrichment, store the collected data in a secure database, and maintain continuity of knowledge on measured cylinders until IAEA inspector arrival. The three main objectives of this FY09 project are summarized here and described in more detail in the report: (1) Develop a preliminary design for a prototype NDA system, (2) Refine PNNL's MCNP models of the NDA system, and (3) Procure and test key pulse-processing components. Progress against these tasks to date, and next steps, are discussed.

  18. Method for isotope enrichment by photoinduced chemiionization

    DOE Patents [OSTI]

    Dubrin, James W.

    1985-01-01

    Isotope enrichment, particularly .sup.235 U enrichment, is achieved by irradiating an isotopically mixed vapor feed with radiant energy at a wavelength or wavelengths chosen to selectively excite the species containing a desired isotope to a predetermined energy level. The vapor feed if simultaneously reacted with an atomic or molecular reactant species capable of preferentially transforming the excited species into an ionic product by a chemiionization reaction. The ionic product, enriched in the desired isotope, is electrostatically or electromagnetically extracted from the reaction system.

  19. Aseismic design criteria for uranium enrichment plants

    SciTech Connect (OSTI)

    Beavers, J.E.

    1980-01-01

    In this paper technological, economical, and safety issues of aseismic design of uranium enrichment plants are presented. The role of management in the decision making process surrounding these issues is also discussed. The resolution of the issues and the decisions made by management are controlling factors in developing aseismic design criteria for any facility. Based on past experience in developing aseismic design criteria for the GCEP various recommendations are made for future enrichment facilities, and since uranium enrichment plants are members of the nuclear fuel cycle the discussion and recommendations presented herein are applicable to other nonreactor nuclear facilities.

  20. Selective enrichment of a methanol-utilizing consortium using pulp & paper mill waste streams

    SciTech Connect (OSTI)

    Gregory R. Mockos; William A. Smith; Frank J. Loge; David N. Thompson

    2007-04-01

    Efficient utilization of carbon inputs is critical to the economic viability of the current forest products sector. Input carbon losses occur in various locations within a pulp mill, including losses as volatile organics and wastewater . Opportunities exist to capture this carbon in the form of value-added products such as biodegradable polymers. Waste activated sludge from a pulp mill wastewater facility was enriched for 80 days for a methanol-utilizing consortium with the goal of using this consortium to produce biopolymers from methanol-rich pulp mill waste streams. Five enrichment conditions were utilized: three high-methanol streams from the kraft mill foul condensate system, one methanol-amended stream from the mill wastewater plant, and one methanol-only enrichment. Enrichment reactors were operated aerobically in sequencing batch mode at neutral pH and 25°C with a hydraulic residence time and a solids retention time of four days. Non-enriched waste activated sludge did not consume methanol or reduce chemical oxygen demand. With enrichment, however, the chemical oxygen demand reduction over 24 hour feed/decant cycles ranged from 79 to 89 %, and methanol concentrations dropped below method detection limits. Neither the non-enriched waste activated sludge nor any of the enrichment cultures accumulated polyhydroxyalkanoates (PHAs) under conditions of nitrogen sufficiency. Similarly, the non-enriched waste activated sludge did not accumulate PHAs under nitrogen limited conditions. By contrast, enriched cultures accumulated PHAs to nearly 14% on a dry weight basis under nitrogen limited conditions. This indicates that selectively-enriched pulp mill waste activated sludge can serve as an inoculum for PHA production from methanol-rich pulp mill effluents.

  1. Safeguards Issues at Nuclear Reactors and Enrichment Plants ...

    Office of Scientific and Technical Information (OSTI)

    Safeguards Issues at Nuclear Reactors and Enrichment Plants Citation Details In-Document Search Title: Safeguards Issues at Nuclear Reactors and Enrichment Plants The Agency's ...

  2. Safeguards Issues at Nuclear Reactors and Enrichment Plants ...

    Office of Scientific and Technical Information (OSTI)

    Safeguards Issues at Nuclear Reactors and Enrichment Plants Citation Details In-Document Search Title: Safeguards Issues at Nuclear Reactors and Enrichment Plants Authors: Boyer, ...

  3. Safeguards at Gas Centrifuge Enrichment Plants: Why is Iran a...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: Safeguards at Gas Centrifuge Enrichment Plants: Why is Iran a Threat? Citation Details In-Document Search Title: Safeguards at Gas Centrifuge Enrichment Plants: ...

  4. Safeguards at Gas Centrifuge Enrichment Plants: Why is Iran a...

    Office of Scientific and Technical Information (OSTI)

    Safeguards at Gas Centrifuge Enrichment Plants: Why is Iran a Threat? Citation Details In-Document Search Title: Safeguards at Gas Centrifuge Enrichment Plants: Why is Iran a ...

  5. News Media invited to interview JLab summer, science enrichment...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    JLab summer, science enrichment program participants; cover closing Poster Session 2003 ... News Media invited to interview JLab summer, science enrichment program participants; ...

  6. Jefferson Lab seeks applicants for summer, science teacher enrichment...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    seeks applicants for summer, science teacher enrichment program Jefferson Lab seeks applicants for summer, science teacher enrichment program February 26, 2003 Calling all middle ...

  7. Oxygen-Enriched Combustion for Military Diesel Engine Generators...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Oxygen-Enriched Combustion for Military Diesel Engine Generators Substantial increases in brake power and considerably lower peak pressure can result from oxygen-enriched diesel ...

  8. Safeguards Issues at Nuclear Reactors and Enrichment Plants ...

    Office of Scientific and Technical Information (OSTI)

    Safeguards Issues at Nuclear Reactors and Enrichment Plants Citation Details In-Document Search Title: Safeguards Issues at Nuclear Reactors and Enrichment Plants You are ...

  9. Toxic Substances Control Act Uranium Enrichment Federal Facility...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance Agreement Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance Agreement Toxic ...

  10. Enrichment Determination of Uranium in Shielded Configurations

    SciTech Connect (OSTI)

    Crye, Jason Michael; Hall, Howard L; McConchie, Seth M; Mihalczo, John T; Pena, Kirsten E

    2011-01-01

    The determination of the enrichment of uranium is required in many safeguards and security applications. Typical methods of determining the enrichment rely on detecting the 186 keV gamma ray emitted by {sup 235}U. In some applications, the uranium is surrounded by external shields, and removal of the shields is undesirable. In these situations, methods relying on the detection of the 186 keV gamma fail because the gamma ray is shielded easily. Oak Ridge National Laboratory (ORNL) has previously measured the enrichment of shielded uranium metal using active neutron interrogation. The method consists of measuring the time distribution of fast neutrons from induced fissions with large plastic scintillator detectors. To determine the enrichment, the measurements are compared to a calibration surface that is created from Monte Carlo simulations where the enrichment in the models is varied. In previous measurements, the geometry was always known. ORNL is extending this method to situations where the geometry and materials present are not known in advance. In the new method, the interrogating neutrons are both time and directionally tagged, and an array of small plastic scintillators measures the uncollided interrogating neutrons. Therefore, the attenuation through the item along many different paths is known. By applying image reconstruction techniques, an image of the item is created which shows the position-dependent attenuation. The image permits estimating the geometry and materials present, and these estimates are used as input for the Monte Carlo simulations. As before, simulations predict the time distribution of induced fission neutrons for different enrichments. Matching the measured time distribution to the closest prediction from the simulations provides an estimate of the enrichment. This presentation discusses the method and provides results from recent simulations that show the importance of knowing the geometry and materials from the imaging system.

  11. Operations Manual for the Portable NDA II Equipment (Version 2.2)

    SciTech Connect (OSTI)

    Bandong, B B; Wong, J L; Valentine, J D; Decman, D J

    2002-06-27

    This document describes the operation and use of the Portable Nondestructive Assay (NDA) II equipment for use in the determination of {sup 235}U enrichment of uranium of various chemical forms and contained in different vessels. The Portable NDA II is the next generation NDA equipment assembled by Lawrence Livermore National Laboratory (LLNL) for the Department of Energy's Highly Enriched Uranium-Transparency Implementation Program (HEU-TIP). Presented in this document is an overview of the enrichment measurement methodology, instructions for the assembly and disassembly of the equipment, description of and user's guide for the UMeter enrichment meter software and a section on system troubleshooting. Also included herewith are facility-specific information and parameters for each of the HEU-processing sites subject to the HEU Transparency Implementation Program.

  12. Recovery and Blend-Down Uranium for Beneficial use in Commercial Reactors - 13373

    SciTech Connect (OSTI)

    Magoulas, Virginia [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)] [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)

    2013-07-01

    In April 2001 the Department of Energy (DOE) and the Tennessee Valley Authority (TVA) signed an Interagency Agreement to transfer approximately 33 MT of off-specification (off-spec) highly enriched uranium (HEU) from DOE to TVA for conversion to commercial reactor fuel. Since that time additional surplus off-spec HEU material has been added to the program, making the total approximately 46 MT off-spec HEU. The disposition path for approximately half (23 MT) of this 46 MT of surplus HEU material, was down blending through the H-canyon facility at the Savannah River Site (SRS). The HEU is purified through the H-canyon processes, and then blended with natural uranium (NU) to form low enriched uranium (LEU) solution with a 4.95% U-235 isotopic content. This material was then transported to a TVA subcontractor who converted the solution to uranium oxide and then fabricated into commercial light water reactor (LWR) fuel. This fuel is now powering TVA reactors and supplying electricity to approximately 1 million households in the TVA region. There is still in excess of approximately 10 to 14 MT of off-spec HEU throughout the DOE complex or future foreign and domestic research reactor returns that could be recovered and down blended for use in either currently designed light water reactors, ?5% enriched LEU, or be made available for use in subsequent advanced 'fast' reactor fuel designs, ?19% LEU. (authors)

  13. Multi-stage combustion using nitrogen-enriched air

    DOE Patents [OSTI]

    Fischer, Larry E.; Anderson, Brian L.

    2004-09-14

    Multi-stage combustion technology combined with nitrogen-enriched air technology for controlling the combustion temperature and products to extend the maintenance and lifetime cycles of materials in contact with combustion products and to reduce pollutants while maintaining relatively high combustion and thermal cycle efficiencies. The first stage of combustion operates fuel rich where most of the heat of combustion is released by burning it with nitrogen-enriched air. Part of the energy in the combustion gases is used to perform work or to provide heat. The cooled combustion gases are reheated by additional stages of combustion until the last stage is at or near stoichiometric conditions. Additional energy is extracted from each stage to result in relatively high thermal cycle efficiency. The air is enriched with nitrogen using air separation technologies such as diffusion, permeable membrane, absorption, and cryogenics. The combustion method is applicable to many types of combustion equipment, including: boilers, burners, turbines, internal combustion engines, and many types of fuel including hydrogen and carbon-based fuels including methane and coal.

  14. Uranium enrichment management review: summary of analysis

    SciTech Connect (OSTI)

    Not Available

    1981-01-01

    In May 1980, the Assistant Secretary for Resource Applications within the Department of Energy requested that a group of experienced business executives be assembled to review the operation, financing, and management of the uranium enrichment enterprise as a basis for advising the Secretary of Energy. After extensive investigation, analysis, and discussion, the review group presented its findings and recommendations in a report on December 2, 1980. The following pages contain background material on which that final report was based. This report is arranged in chapters that parallel those of the uranium enrichment management review final report - chapters that contain summaries of the review group's discussion and analyses in six areas: management of operations and construction; long-range planning; marketing of enrichment services; financial management; research and development; and general management. Further information, in-depth analysis, and discussion of suggested alternative management practices are provided in five appendices.

  15. Status of gadolinium enrichment technology at LLNL

    SciTech Connect (OSTI)

    Haynam, C.; Comaskey, B.; Conway, J.; Eggert, J.; Glaser, J.; Ng, E.; Paisner, J.; Solarz, R.; Worden, E.

    1993-01-01

    A method based on,polarization selectivity and three step laser photoionization is presented for separation of the odd isotopes of gadolinium. Measurements of the spectroscopic parameters needed to quantify the excitation pathway are discussed. Model results are presented for the efficiency of photoionization. The vapor properties of electron beam vaporized gadolinium are presented which show dramatic cooling during the expansion of the hot dense vapor into a vacuum. This results in a significant increase in the efficiency of conversion of natural feed into enriched product in the AVLIS process. Production of enriched gadolinium for use in commercial power reactors appears to be economically viable using technology in use at LLNL.

  16. Onsite Gaseous Centrifuge Enrichment Plant UF6 Cylinder Destructive Analysis

    SciTech Connect (OSTI)

    Anheier, Norman C.; Cannon, Bret D.; Qiao, Hong; Carter, Jennifer C.; McNamara, Bruce K.; O'Hara, Matthew J.; Phillips, Jon R.; Curtis, Michael M.

    2012-07-17

    The IAEA safeguards approach for gaseous centrifuge enrichment plants (GCEPs) includes measurements of gross, partial, and bias defects in a statistical sampling plan. These safeguard methods consist principally of mass and enrichment nondestructive assay (NDA) verification. Destructive assay (DA) samples are collected from a limited number of cylinders for high precision offsite mass spectrometer analysis. DA is typically used to quantify bias defects in the GCEP material balance. Under current safeguards measures, the operator collects a DA sample from a sample tap following homogenization. The sample is collected in a small UF6 sample bottle, then sealed and shipped under IAEA chain of custody to an offsite analytical laboratory. Current practice is expensive and resource intensive. We propose a new and novel approach for performing onsite gaseous UF6 DA analysis that provides rapid and accurate assessment of enrichment bias defects. DA samples are collected using a custom sampling device attached to a conventional sample tap. A few micrograms of gaseous UF6 is chemically adsorbed onto a sampling coupon in a matter of minutes. The collected DA sample is then analyzed onsite using Laser Ablation Absorption Ratio Spectrometry-Destructive Assay (LAARS-DA). DA results are determined in a matter of minutes at sufficient accuracy to support reliable bias defect conclusions, while greatly reducing DA sample volume, analysis time, and cost.

  17. Low temperature combustion using nitrogen enrichment to mitigate NOx from large bore natural gas fueled engines.

    SciTech Connect (OSTI)

    Biruduganti, M.; Gupta, S.; Sekar, R.; Energy Systems

    2010-01-01

    Low temperature combustion is identified as one of the pathways to meet the mandatory ultra low NO{sub x} emissions levels set by the regulatory agencies. Exhaust gas recirculation (EGR) is a well known technique to realize low NO{sub x} emissions. However, EGR has many built-in adverse ramifications that negate its advantages in the long term. This paper discusses nitrogen enrichment of intake air using air separation membranes as a better alternative to the mature EGR technique. This investigation was undertaken to determine the maximum acceptable level of nitrogen enrichment of air for a single-cylinder spark-ignited natural gas engine. NO{sub x} reduction as high as 70% was realized with a modest 2% nitrogen enrichment while maintaining power density and simultaneously improving fuel conversion efficiency (FCE). Any enrichment beyond this level degraded engine performance in terms of power density, FCE, and unburned hydrocarbon emissions. The effect of ignition timing was also studied with and without N{sub 2} enrichment. Finally, lean burn versus stoichiometric operation utilizing nitrogen enrichment was compared. Analysis showed that lean burn operation along with nitrogen enrichment is one of the effective pathways for realizing better FCE and lower NO{sub x} emissions.

  18. Benchmark Evaluation of Uranium Metal Annuli and Cylinders with Beryllium Reflectors

    SciTech Connect (OSTI)

    John D. Bess

    2010-06-01

    An extensive series of delayed critical experiments were performed at the Oak Ridge Critical Experiments Facility using enriched uranium metal during the 1960s and 1970s in support of criticality safety operations at the Y-12 Plant. These experiments were designed to evaluate the storage, casting, and handling limits of the Y-12 Plant and to provide data for the verification of cross sections and calculation methods utilized in nuclear criticality safety applications. Many of these experiments have already been evaluated and included in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook: unreflected (HEU-MET-FAST-051), graphite-reflected (HEU-MET-FAST-071), and polyethylene-reflected (HEU-MET-FAST-076). Three of the experiments consisted of highly-enriched uranium (HEU, ~93.2% 235U) metal parts reflected by beryllium metal discs. The first evaluated experiment was constructed from a stack of 7-in.-diameter, 4-1/8-in.-high stack of HEU discs top-reflected by a 7-in.-diameter, 5-9/16-in.-high stack of beryllium discs. The other two experiments were formed from stacks of concentric HEU metal annular rings surrounding a 7-in.diameter beryllium core. The nominal outer diameters were 13 and 15 in. with a nominal stack height of 5 and 4 in., respectively. These experiments have been evaluated for inclusion in the ICSBEP Handbook.

  19. Gas Centrifuge Enrichment Plant Safeguards System Modeling

    SciTech Connect (OSTI)

    Elayat, H A; O'Connell, W J; Boyer, B D

    2006-06-05

    The U.S. Department of Energy (DOE) is interested in developing tools and methods for potential U.S. use in designing and evaluating safeguards systems used in enrichment facilities. This research focuses on analyzing the effectiveness of the safeguards in protecting against the range of safeguards concerns for enrichment plants, including diversion of attractive material and unauthorized modes of use. We developed an Extend simulation model for a generic medium-sized centrifuge enrichment plant. We modeled the material flow in normal operation, plant operational upset modes, and selected diversion scenarios, for selected safeguards systems. Simulation modeling is used to analyze both authorized and unauthorized use of a plant and the flow of safeguards information. Simulation tracks the movement of materials and isotopes, identifies the signatures of unauthorized use, tracks the flow and compilation of safeguards data, and evaluates the effectiveness of the safeguards system in detecting misuse signatures. The simulation model developed could be of use to the International Atomic Energy Agency IAEA, enabling the IAEA to observe and draw conclusions that uranium enrichment facilities are being used only within authorized limits for peaceful uses of nuclear energy. It will evaluate improved approaches to nonproliferation concerns, facilitating deployment of enhanced and cost-effective safeguards systems for an important part of the nuclear power fuel cycle.

  20. Method for laser induced isotope enrichment

    DOE Patents [OSTI]

    Pronko, Peter P.; Vanrompay, Paul A.; Zhang, Zhiyu

    2004-09-07

    Methods for separating isotopes or chemical species of an element and causing enrichment of a desired isotope or chemical species of an element utilizing laser ablation plasmas to modify or fabricate a material containing such isotopes or chemical species are provided. This invention may be used for a wide variety of materials which contain elements having different isotopes or chemical species.

  1. USE OF MAILBOX APPROACH, VIDEO SURVEILLANCE, AND SHORT-NOTICE RANDOM INSPECTIONS TO ENHANCE DETECTION OF UNDECLARED LEU PRODUCTION AT GAS CENTRIFUGE ENRICHMENT PLANTS.

    SciTech Connect (OSTI)

    BOYER, B.D.; GORDON, D.M.; JO, J.

    2006-07-16

    Current safeguards approaches used by the IAEA at gas centrifuge enrichment plants (GCEPs) need enhancement in order to detect undeclared LEU production with adequate detection probability. ''Mailbox'' declarations have been used in the last two decades to verify receipts, production, and shipments at some bulk-handling facilities (e.g., fuel-fabrication plants). The operator declares the status of his plant to the IAEA on a daily basis using a secure ''Mailbox'' system such as a secure tamper-resistant computer. The operator agrees to hold receipts and shipments for a specified period of time, along with a specified number of annual inspections, to enable inspector access to a statistically large enough population of UF{sub 6} cylinders and fuel assemblies to achieve the desired detection probability. The inspectors can access the ''Mailbox'' during randomly timed inspections and then verify the operator's declarations for that day. Previously, this type of inspection regime was considered mainly for verifying the material balance at fuel-fabrication, enrichment, and conversion plants. Brookhaven National Laboratory has expanded the ''Mailbox'' concept with short-notice random inspections (SNRIs), coupled with enhanced video surveillance, to include declaration and verification of UF{sub 6} cylinder operational data to detect activities associated with undeclared LEU production at GCEPs. Since the ''Mailbox'' declarations would also include data relevant to material-balance verification, these randomized inspections would replace the scheduled monthly interim inspections for material-balance purposes; in addition, the inspectors could simultaneously perform the required number of Limited-Frequency Unannounced Access (LFUA) inspections used for HEU detection. This approach would provide improved detection capabilities for a wider range of diversion activities with not much more inspection effort than at present.

  2. US developments in technology for uranium enrichment

    SciTech Connect (OSTI)

    Wilcox, W.J. Jr.; McGill, R.M.

    1982-01-01

    The purpose of this paper is to review recent progress and the status of the work in the United States on that part of the fuel cycle concerned with uranium enrichment. The United States has one enrichment process, gaseous diffusion, which has been continuously operated in large-scale production for the past 37 years; another process, gas centrifugation, which is now in the construction phase; and three new processes, molecular laser isotope separation, atomic vapor laser isotope separation, plasma separation process, in which the US has also invested sizable research and development efforts over the last few years. The emphasis in this paper is on the technical aspects of the various processes, but the important economic factors which will define the technological mix which may be applied in the next two decades are also discussed.

  3. Uranium Mining, Conversion, and Enrichment Industries

    Energy Savers [EERE]

    i Analysis of Potential Impacts of Uranium Transfers on the Domestic Uranium Mining, Conversion, and Enrichment Industries May 1, 2015 ii EXECUTIVE SUMMARY: The Department of Energy ("Department" or "DOE") plans to transfer the equivalent of up to 2,100 metric tons ("MTU") of natural uranium per year (with a higher total for calendar year 2015, mainly because of transfers already executed or under way before today's determination). These transfers would include

  4. EIS-0471: Areva Eagle Rock Enrichment Facility in Bonneville...

    Office of Environmental Management (EM)

    1: Areva Eagle Rock Enrichment Facility in Bonneville County, ID EIS-0471: Areva Eagle Rock Enrichment Facility in Bonneville County, ID May 20, 2011 delete me old download page ...

  5. Report of Survey of Oak Ridge Isotope Enrichment (Calutron) Facility...

    Office of Environmental Management (EM)

    Report of Survey of Oak Ridge Isotope Enrichment (Calutron) Facility Building 9204-3 Report of Survey of Oak Ridge Isotope Enrichment (Calutron) Facility Building 9204-3 The ...

  6. Study of using oxygen-enriched combustion air for locomotive diesel engines

    SciTech Connect (OSTI)

    Poola, R.B.; Sekar, R.; Assanis, D.N.; Cataldi, G.R.

    1996-12-31

    A thermodynamic simulation is used to study the effects of oxygen-enriched intake air on the performance and nitrogen oxide (NO) emissions of a locomotive diesel engine. The parasitic power of the air separation membrane required to supply the oxygen-enriched air is also estimated. For a given constraint on peak cylinder pressure, the gross and net power output of an engine operating under different levels of oxygen enrichment are compared with those obtained when a high-boost turbocharged engine is used. A 4% increase in peak cylinder pressure can result in an increase in net engine power of approximately 13% when intake air with an oxygen content of 28% by volume is used and fuel injection timing is retarded by 4 degrees. When the engine is turbocharged to a higher inlet boost, the same increase in peak cylinder pressure can improve power by only 4%. If part of the significantly higher exhaust enthalpies available as a result of oxygen enrichment are recovered, the power requirements of the air separator membrane can be met, resulting in substantial net power improvements. Oxygen enrichment with its attendant higher combustion temperatures, reduces emissions of particulates and visible smoke but increases NO emissions (by up to three times at 26% oxygen content). Therefore, exhaust gas after-treatment and heat recovery would be required if the full potential of oxygen enrichment for improving the performance of locomotive diesel engines is to be realized.

  7. DOE Signs Advanced Enrichment Technology License and Facility Lease |

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Department of Energy Advanced Enrichment Technology License and Facility Lease DOE Signs Advanced Enrichment Technology License and Facility Lease December 8, 2006 - 9:34am Addthis Announces Agreements with USEC Enabling Deployment of Advanced Domestic Technology for Uranium Enrichment WASHINGTON, DC - U.S. Secretary of Energy Samuel W. Bodman today announced the signing of a lease agreement with the United States Enrichment Corporation, Inc. (USEC) for their use of the Department's gas

  8. Oxygen-Enriched Combustion for Military Diesel Engine Generators

    Broader source: Energy.gov [DOE]

    Substantial increases in brake power and considerably lower peak pressure can result from oxygen-enriched diesel combustion

  9. GLOBAL THREAT REDUCTION INITIATIVE REACTOR CONVERSION PROGRAM: STATUS AND CURRENT PLANS

    SciTech Connect (OSTI)

    Staples, Parrish A.; Leach, Wayne; Lacey, Jennifer M.

    2009-10-07

    The U.S. Department of Energys National Nuclear Security Administration (NNSA) Reactor Conversion Program supports the minimization, and to the extent possible, elimination of the use of high enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors and radioisotope production processes to the use of low enriched uranium (LEU). The Reactor Conversion Program is a technical pillar of the NNSA Global Threat Reduction Initiative (GTRI) which is a key organization for implementing U.S. HEU minimization policy and works to reduce and protect vulnerable nuclear and radiological material domestically and abroad.

  10. Chapter 20 - Uranium Enrichment Decontamination & Decommissioning Fund

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    0. Uranium Enrichment Decontamination and Decommissioning Fund 20-1 CHAPTER 20 URANIUM ENRICHMENT DECONTAMINATION AND DECOMMISSIONING FUND 1. INTRODUCTION. a. Purpose. To establish policies and procedures for the financial management, accounting, budget preparation, cash management of the Uranium Enrichment Decontamination and Decommissioning Fund, referred to hereafter as the Fund. b. Applicability. This chapter applies to all Departmental elements, including the National Nuclear Security

  11. Supplemental Reactor Physics Calculations and Analysis of ELF Mk 1A Fuel

    SciTech Connect (OSTI)

    Michael A. Pope

    2014-10-01

    These calculations supplement previous the reactor physics work evaluating the Enhanced Low Enriched Uranium (LEU) Fuel (ELF) Mk 1A element. This includes various additional comparisons between the current Highly Enriched Uranium (HEU) and LEU along with further characterization of the performance of the ELF fuel. The excess reactivity to be held down at BOC for ELF Mk 1A fuel is estimated to be approximately $2.75 greater than with HEU for a typical cycle. This is a combined effect of the absence of burnable poison in the ELF fuel and the reduced neck shim worth in LEU fuel compared to HEU. Burnable poison rods were conceptualized for use in the small B positions containing Gd2O3 absorber. These were shown to provide $2.37 of negative reactivity at BOC and to burn out in less than half of a cycle. The worth of OSCCs is approximately the same between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. This was evaluated by rotating all banks simultaneously. The safety rod worth is relatively unchanged between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. However, this should be reevaluated with different loadings. Neutron flux, both total and fast (>1 MeV), is either the same or reduced upon changing from HEU to ELF Mk 1A (LEU) fuels in the representative loading evaluated. This is consistent with the well-established trend of lower neutron fluxes for a given power in LEU than HEU.The IPT loop void reactivity is approximately the same or less positive with ELF Mk 1A (LEU) fuel than HEU in the representative loading evaluated.

  12. Utilization of non-weapons-grade plutonium and highly enriched...

    Office of Scientific and Technical Information (OSTI)

    the VVER reactors using thorium and heavy water Citation Details In-Document Search Title: ... the VVER reactors using thorium and heavy water A method for joint utilization of ...

  13. Belgium Highly Enriched Uranium and Plutonium Removals | National...

    National Nuclear Security Administration (NNSA)

    This Site Budget IG Web Policy Privacy No Fear Act Accessibility FOIA Sitemap Federal Government The White House DOE.gov USA.gov Jobs Apply for Our Jobs Our Jobs Working at NNSA...

  14. GTRI's Convert Program: Minimizing the Use of Highly Enriched...

    National Nuclear Security Administration (NNSA)

    This Site Budget IG Web Policy Privacy No Fear Act Accessibility FOIA Sitemap Federal Government The White House DOE.gov USA.gov Jobs Apply for Our Jobs Our Jobs Working at NNSA...

  15. NNSA Authorizes Start-Up of Highly Enriched Uranium Materials...

    National Nuclear Security Administration (NNSA)

    This Site Budget IG Web Policy Privacy No Fear Act Accessibility FOIA Sitemap Federal Government The White House DOE.gov USA.gov Jobs Apply for Our Jobs Our Jobs Working at NNSA...

  16. Quadrilateral Cooperation on High-density Low-enriched Uranium...

    National Nuclear Security Administration (NNSA)

    This Site Budget IG Web Policy Privacy No Fear Act Accessibility FOIA Sitemap Federal Government The White House DOE.gov USA.gov Jobs Apply for Our Jobs Our Jobs Working at NNSA...

  17. Italy Highly Enriched Uranium and Plutonium Removals | National...

    National Nuclear Security Administration (NNSA)

    This Site Budget IG Web Policy Privacy No Fear Act Accessibility FOIA Sitemap Federal Government The White House DOE.gov USA.gov Jobs Apply for Our Jobs Our Jobs Working at NNSA...

  18. NNSA Highly Enriched Uranium Removal Featured on The Rachel Maddow...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    NNSA. Michael Hess Michael Hess Former Digital Communications Specialist, Office of ... the news with Rachel Maddow. The full video can be found on the original NNSA blog post. ...

  19. US, Kazakhstan Cooperate to Eliminate Highly Enriched Uranium...

    National Nuclear Security Administration (NNSA)

    ... application of nuclear science. NNSA maintains and enhances the safety, security, reliability and performance of the U.S. nuclear weapons stockpile without nuclear testing; works ...

  20. Application of oxygen-enriched combustion for locomotive diesel engines. Phase 1

    SciTech Connect (OSTI)

    Poola, R.B.; Sekar, R.R.; Assanis, D.N.

    1996-09-01

    A thermodynamic simulation is used to study the effects of oxygen-enriched intake air on the performance and nitrogen oxide (NO) emissions of a locomotive diesel engine. The parasitic power of the air separation membrane required to supply the oxygen-enriched air is also estimated. For a given constraint on peak cylinder pressure, the gross and net power outputs of an engine operating under different levels of oxygen enrichment are compared with those obtained when a high-boost turbocharged engine is used. A 4% increase in peak cylinder pressure can result in an increase in net engine power of approximately 13% when intake air with an oxygen content of 28% by volume is used and fuel injection timing is retarded by 4 degrees. When the engine is turbocharged to a higher inlet boost, the same increase in peak cylinder pressure improves power by only 4%. If part of the significantly higher exhaust enthalpies available as a result of oxygen enrichment are recovered, the power requirements of the air separator membrane can be met, resulting in substantial net power improvements. Oxygen enrichment reduces particulate and visible smoke emissions but increases NO emissions. However, a combination of retarded fuel injection timing and post-treatment of exhaust gases may be adequate to meet the locomotive diesel engine NO{sub x} standards. Exhaust gas after-treatment and heat recovery would be required to realize the full potential of oxygen enrichment. Economic analysis shows that oxygen-enrichment technology is economically feasible and provides high returns on investment. The study also indicates the strong influence of membrane parasitic requirements and exhaust energy recovery on economic benefits. To obtain an economic advantage while using a membrane with higher parasitic power requirements, it is necessary to recover a part of the exhaust energy.

  1. Safeguarding a NWS International Enrichment Center as an Enriched Uranium Store

    SciTech Connect (OSTI)

    Curtis, Michael M.

    2008-03-31

    The operational and regulatory singularities of a multilateral facility designed to provide enriched uranium to a consortium of members may engender a new sub-category of safeguard criteria for the International Atomic Energy Agency (IAEA). This paper introduces the contingency of monitoring such a facility as a uranium storage center with cylinders containing low-enriched uranium (LEU) as the principal, and perhaps only, material open to verification. Accountancy and verification techniques will be proffered together with disparate means for maintaining continuity of knowledge (CoK) on verified stock.

  2. NNSA Administrator Klotz Tours Argonne National Laboratory in...

    National Nuclear Security Administration (NNSA)

    from the use of HEU-which can be a proliferation risk-to low enriched uranium (LEU) fuel. ... (Mo-99), without the use of proliferation-sensitive HEU. Argonne supports ...

  3. Uranium enrichment. Printed at the request of the Committee on Energy and Natural Resources, United States Senate, May 1982

    SciTech Connect (OSTI)

    Not Available

    1982-01-01

    Two congressional reports outline the need for new uranium-enrichment plants and their costs. Part I, The Need for Additional Uranium Enrichment Capacity to Meet Demand, examines DOE's case for continuing construction of the Portsmouth, Ohio gas centrifuge plant on the basis of projected demand. The report concludes that DOE projections are high and that future demand can be met through preproduction and stockpiling. Part II, Necessity for GCEP (Gas Centrifuge Enrichment Plant) Under Low Nuclear Power Growth Conditions, concludes that continued construction is economically valid because of the uncertainty of demand forecasts. 79 references, 12 tables. (DCK)

  4. Technical solutions to nonproliferation challenges

    SciTech Connect (OSTI)

    Satkowiak, Lawrence

    2014-05-09

    The threat of nuclear terrorism is real and poses a significant challenge to both U.S. and global security. For terrorists, the challenge is not so much the actual design of an improvised nuclear device (IND) but more the acquisition of the special nuclear material (SNM), either highly enriched uranium (HEU) or plutonium, to make the fission weapon. This paper provides two examples of technical solutions that were developed in support of the nonproliferation objective of reducing the opportunity for acquisition of HEU. The first example reviews technologies used to monitor centrifuge enrichment plants to determine if there is any diversion of uranium materials or misuse of facilities to produce undeclared product. The discussion begins with a brief overview of the basics of uranium processing and enrichment. The role of the International Atomic Energy Agency (IAEA), its safeguard objectives and how the technology evolved to meet those objectives will be described. The second example focuses on technologies developed and deployed to monitor the blend down of 500 metric tons of HEU from Russia's dismantled nuclear weapons to reactor fuel or low enriched uranium (LEU) under the U.S.-Russia HEU Purchase Agreement. This reactor fuel was then purchased by U.S. fuel fabricators and provided about half the fuel for the domestic power reactors. The Department of Energy established the HEU Transparency Program to provide confidence that weapons usable HEU was being blended down and thus removed from any potential theft scenario. Two measurement technologies, an enrichment meter and a flow monitor, were combined into an automated blend down monitoring system (BDMS) and were deployed to four sites in Russia to provide 24/7 monitoring of the blend down. Data was downloaded and analyzed periodically by inspectors to provide the assurances required.

  5. Gas centrifuge enrichment plants inspection frequency and remote monitoring issues for advanced safeguards implementation

    SciTech Connect (OSTI)

    Boyer, Brian David; Erpenbeck, Heather H; Miller, Karen A; Ianakiev, Kiril D; Reimold, Benjamin A; Ward, Steven L; Howell, John

    2010-09-13

    Current safeguards approaches used by the IAEA at gas centrifuge enrichment plants (GCEPs) need enhancement in order to verify declared low enriched uranium (LEU) production, detect undeclared LEU production and detect high enriched uranium (BEU) production with adequate probability using non destructive assay (NDA) techniques. At present inspectors use attended systems, systems needing the presence of an inspector for operation, during inspections to verify the mass and {sup 235}U enrichment of declared cylinders of uranium hexafluoride that are used in the process of enrichment at GCEPs. This paper contains an analysis of how possible improvements in unattended and attended NDA systems including process monitoring and possible on-site destructive analysis (DA) of samples could reduce the uncertainty of the inspector's measurements providing more effective and efficient IAEA GCEPs safeguards. We have also studied a few advanced safeguards systems that could be assembled for unattended operation and the level of performance needed from these systems to provide more effective safeguards. The analysis also considers how short notice random inspections, unannounced inspections (UIs), and the concept of information-driven inspections can affect probability of detection of the diversion of nuclear material when coupled to new GCEPs safeguards regimes augmented with unattended systems. We also explore the effects of system failures and operator tampering on meeting safeguards goals for quantity and timeliness and the measures needed to recover from such failures and anomalies.

  6. Ultraheavy element enrichment in impulsive solar flares

    SciTech Connect (OSTI)

    Eichler, David

    2014-10-10

    Particle acceleration by cascading Alfvén wave turbulence was suggested as being responsible for energetic particle populations in {sup 3}He-rich solar flares. In particular, it was noted that the damping of the turbulence by the tail of the particle distribution in rigidity naturally leads to the dramatic enhancement of a pre-accelerated species—as {sup 3}He is posited to be—and superheavy elements. The subsequent detection of large enrichment of ultraheavies, relative to iron, has apparently confirmed this prediction, lending support to the original idea. It is shown here that this picture could be somewhat sharpened by progress in understanding the three-dimensional geometrical details of cascading Alfvén turbulence. The mechanism may be relevant in other astrophysical environments where the source of turbulence is nonmagnetic, such as clusters of galaxies.

  7. High-temperature gas-cooled reactors: preliminary safety and environmental information document. Volume IV

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

    Information is presented concerning medium-enriched uranium/thorium once-through fuel cycle; medium-enrichment uranium-233/thorium recycle fuel; high-enrichment uranium-235/thorium recycle (spiked) fuel cycle; high-enrichment uranium-233/thorium recycle (spiked) fuel cycle; and gas-turbine high-temperature gas-cooled reactor.

  8. Analysis of Enriched Uranyl Nitrate in Nested Annular Tank Array

    SciTech Connect (OSTI)

    John D. Bess; James D. Cleaver

    2009-06-01

    Two series of experiments were performed at the Rocky Flats Critical Mass Laboratory during the 1980s using highly enriched (93%) uranyl nitrate solution in annular tanks. [1, 2] Tanks were of typical sizes found in nuclear production plants. Experiments looked at tanks of varying radii in a co-located set of nested tanks, a 1 by 2 array, and a 1 by 3 array. The co-located set of tanks had been analyzed previously [3] as a benchmark for inclusion within the International Handbook of Evaluated Criticality Safety Benchmark Experiments. [4] The current study represents the benchmark analysis of the 1 by 3 array of a series of nested annular tanks. Of the seventeen configurations performed in this set of experiments, twelve were evaluated and nine were judged as acceptable benchmarks.

  9. Method and apparatus for measuring enrichment of UF6

    DOE Patents [OSTI]

    Hill, Thomas Roy; Ianakiev, Kiril Dimitrov

    2011-06-07

    A system and method are disclosed for determining the enrichment of .sup.235U in Uranium Hexafluoride (UF6) utilizing synthesized X-rays which are directed at a container test zone containing a sample of UF6. A detector placed behind the container test zone then detects and counts the X-rays which pass through the container and the UF6. In order to determine the portion of the attenuation due to the UF6 gas alone, this count rate may then be compared to a calibration count rate of X-rays passing through a calibration test zone which contains a vacuum, the test zone having experienced substantially similar environmental conditions as the actual test zone. Alternatively, X-rays of two differing energy levels may be alternately directed at the container, where either the container or the UF6 has a high sensitivity to the difference in the energy levels, and the other having a low sensitivity.

  10. Development of uranium metal targets for {sup 99}Mo production

    SciTech Connect (OSTI)

    Wiencek, T.C.; Hofman, G.L.

    1993-10-01

    A substantial amount of high enriched uranium (HEU) is used for the production of medical-grade {sup 99}Mo. Promising methods of producing irradiation targets are being developed and may lead to the reduction or elimination of this HEU use. To substitute low enriched uranium (LEU) for HEU in the production of {sup 99}Mo, the target material may be changed to uranium metal foil. Methods of fabrication are being developed to simplify assembly and disassembly of the targets. Removal of the uranium foil after irradiation without dissolution of the cladding is a primary goal in order to reduce the amount of liquid radioactive waste material produced in the process. Proof-of-concept targets have been fabricated. Destructive testing indicates that acceptable contact between the uranium foil and the cladding can be achieved. Thermal annealing tests, which simulate the cladding/uranium diffusion conditions during irradiation, are underway. Plans are being made to irradiate test targets.

  11. News Media invited to interview JLab summer, science enrichment program

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    participants; cover closing Poster Session | Jefferson Lab JLab summer, science enrichment program participants; cover closing Poster Session 2003 Education Poster Session 2003 Education Poster Session News Media invited to interview JLab summer, science enrichment program participants; cover closing Poster Session July 28, 2004 Newport News, VA. - News Media representatives are invited to interview, photograph or film participants of Jefferson Lab's summer, science enrichment programs as

  12. New Measures to Safeguard Gas Centrifuge Enrichment Plants (Conference) |

    Office of Scientific and Technical Information (OSTI)

    SciTech Connect SciTech Connect Search Results Conference: New Measures to Safeguard Gas Centrifuge Enrichment Plants Citation Details In-Document Search Title: New Measures to Safeguard Gas Centrifuge Enrichment Plants As Gas Centrifuge Enrichment Plants (GCEPs) increase in separative work unit (SWU) capacity, the current International Atomic Energy Agency (IAEA) model safeguards approach needs to be strengthened. New measures to increase the effectiveness of the safeguards approach are

  13. The Enriched Xenon Observatory (Journal Article) | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    The Enriched Xenon Observatory Citation Details In-Document Search Title: The Enriched Xenon Observatory The Enriched Xenon Observatory (EXO) experiment will search for neutrinoless double beta decay of {sup 136}Xe. The EXO Collaboration is actively pursuing both liquid-phase and gas-phase Xe detector technologies with scalability to the ton-scale. The search for neutrinoless double beta decay of {sup 136}Xe is especially attractive because of the possibility of tagging the resulting Ba daughter

  14. Energy Department Selects Global Laser Enrichment for Future...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    ... Energy Department Selects Global Laser Enrichment for Future Operations at Paducah Site Pamela Thompson is retiring from her 37-year federal career. Retiring Procurement Official ...

  15. Toxic Substances Control Act Uranium Enrichment Federal Facilities...

    Office of Environmental Management (EM)

    Toxic Substance Control Act Uranium Enrichment Federal Facilities Compliance Agreement (TSCA-UE- FFCA), February 20, 1992 State Ohio Agreement Type Compliance Agreement Legal ...

  16. Process for preparing a chemical compound enriched in isotope content. [nitrogen 15-enriched nitric acid

    DOE Patents [OSTI]

    Michaels, E.D.

    1981-02-25

    A process to prepare a chemical enriched in isotope content includes: a chemical exchange reaction between a first and second compound which yields an isotopically enriched first compound and an isotopically depleted second compound; the removal of a portion of the first compound as product and the removal of a portion of the second compound as spent material; the conversion of the remainder of the first compound to the second compound for reflux at the product end of the chemical exchange reaction region; the conversion of the remainder of the second compound to the first compound for reflux at the spent material end of the chemical exchange region; and the cycling of the additional chemicals produced by one conversion reaction to the other conversion reaction, for consumption therein. One of the conversion reactions is an oxidation reaction, and the energy that it yields is used to drive the other conversion reaction, a reduction. The reduction reaction is carried out in a solid polymer electrolyte electrolytic reactor. The overall process is energy efficient and yields no waste by-products. A particular embodiment of the process in the production of nitrogen-15-enriched nitric acid.

  17. Hybrid Enrichment Assay Methods for a UF6 Cylinder Verification Station: FY10 Progress Report

    SciTech Connect (OSTI)

    Smith, Leon E.; Jordan, David V.; Orton, Christopher R.; Misner, Alex C.; Mace, Emily K.

    2010-08-01

    Pacific Northwest National Laboratory (PNNL) is developing the concept of an automated UF6 cylinder verification station that would be located at key measurement points to positively identify each cylinder, measure its mass and enrichment, store the collected data in a secure database, and maintain continuity of knowledge on measured cylinders until the arrival of International Atomic Energy Agency (IAEA) inspectors. At the center of this unattended system is a hybrid enrichment assay technique that combines the traditional enrichment-meter method (based on the 186 keV peak from 235U) with non-traditional neutron-induced high-energy gamma-ray signatures (spawned primarily by 234U alpha emissions and 19F(alpha, neutron) reactions). Previous work by PNNL provided proof-of-principle for the non-traditional signatures to support accurate, full-volume interrogation of the cylinder enrichment, thereby reducing the systematic uncertainties in enrichment assay due to UF6 heterogeneity and providing greater sensitivity to material substitution scenarios. The work described here builds on that preliminary evaluation of the non-traditional signatures, but focuses on a prototype field system utilizing NaI(Tl) and LaBr3(Ce) spectrometers, and enrichment analysis algorithms that integrate the traditional and non-traditional signatures. Results for the assay of Type-30B cylinders ranging from 0.2 to 4.95 wt% 235U, at an AREVA fuel fabrication plant in Richland, WA, are described for the following enrichment analysis methods: 1) traditional enrichment meter signature (186 keV peak) as calculated using a square-wave convolute (SWC) algorithm; 2) non-traditional high-energy gamma-ray signature that provides neutron detection without neutron detectors and 3) hybrid algorithm that merges the traditional and non-traditional signatures. Uncertainties for each method, relative to the declared enrichment for each cylinder, are calculated and compared to the uncertainties from an attended HPGe verification station at AREVA, and the IAEA’s uncertainty target values for feed, tail and product cylinders. A summary of the major findings from the field measurements and subsequent analysis follows: • Traditional enrichment-meter assay using specially collimated NaI spectrometers and a Square-Wave-Convolute algorithm can achieve uncertainties comparable to HPGe and LaBr for product, natural and depleted cylinders. • Non-traditional signatures measured using NaI spectrometers enable interrogation of the entire cylinder volume and accurate measurement of absolute 235U mass in product, natural and depleted cylinders. • A hybrid enrichment assay method can achieve lower uncertainties than either the traditional or non-traditional methods acting independently because there is a low degree of correlation in the systematic errors of the two individual methods (wall thickness variation and 234U/235U variation, respectively). This work has indicated that the hybrid NDA method has the potential to serve as the foundation for an unattended cylinder verification station. When compared to today’s handheld cylinder-verification approach, such a station would have the following advantages: 1) improved enrichment assay accuracy for product, tail and feed cylinders; 2) full-volume assay of absolute 235U mass; 3) assay of minor isotopes (234U and 232U) important to verification of feedstock origin; single instrumentation design for both Type 30B and Type 48 cylinders; and 4) substantial reduction in the inspector manpower associated with cylinder verification.

  18. Determination of the 235U Mass and Enrichment within Small UF6 Cylinders via a Neutron Coincidence Well Counting System

    SciTech Connect (OSTI)

    McElroy, Robert Dennis; Croft, Dr. Stephen; Young, Brian M; Venkataraman, Ram

    2011-01-01

    The construction of three new uranium enrichment facilities in the United States has sparked renewed interest in the development and enhancement of methods to determine the enrichment and fissile mass content of UF6 cylinders. We describe the design and examine the expected performance of a UF6 bottle counter developed for the assay of Type 5A cylinders. The counter, as designed and subsequently constructed, is a tall passive neutron well counter with a clam-shell configuration and graphite end plugs operated in fast neutron mode. Factory performance against expectation is described. The relatively high detection efficiency and effectively 4 detection geometry provide a near-ideal measurement configuration, making the UF6 bottle counter a valuable tool for the evaluation of the neutron coincidence approach to UF6 cylinder assay. The impacts of non-uniform filling, voids, enrichment, and mixed enrichments are examined

  19. Safeguards training course: Nuclear material safeguards for enrichment plants

    SciTech Connect (OSTI)

    Not Available

    1990-06-01

    The main objective of this course is to provide the course participants with the necessary skills to perform their inspection activities at enrichment plants. As background information, a variety of enrichment technologies will first be characterized and compared followed by a review of basic cascade, gas centrifuge, and gaseous diffusion theory. To focus on gas centrifuge and gaseous diffusion technology, the major components and system of gas centrifuge and gaseous diffusion enrichment plants including their function in routine LEU production will be identified. The objectives of safeguards at an enrichment plant, including those agreed to in the Hexapartite Safeguards Project, will then be described. Discussions will then focus on potential diversion scenarios at both a centrifuge and diffusion enrichment facility and applicable safeguards inspection activities for detecting these scenarios. This report presents a discussion on basic separation and cascade theory, uranium hexafluoride, and detailed separation theory, including gas centrifuge and gaseous diffusion.

  20. Safeguards training course: Nuclear material safeguards for enrichment plants

    SciTech Connect (OSTI)

    Not Available

    1990-06-01

    The main objective of this training course is to provide the course participants with the necessary skills to perform their inspection activities at enrichment plants. As background information, a variety of enrichment technologies will first be characterized and compared followed by a review of basic cascade, gas centrifuge, and gaseous diffusion theory. To focus on gas centrifuge and gaseous diffusion technology, the major components and systems of gas centrifuge and gaseous diffusion enrichment plants including their function in routine LEU production will be identified. The objectives of safeguards at an enrichment plant, including those agreed to in the Hexapartite Safeguards Project, will then be described. Discussion will then focus on potential diversion scenarios at both a centrifuge and diffusion enrichment facility and applicable safeguards inspection activities for detecting these scenarios.

  1. AISI/DOE Technology Roadmap Program: Development of an O2-Enriched Furnace System for Reduced CO2 and NOx Emissions For the Steel Industry

    SciTech Connect (OSTI)

    Edward W. Grandmaison; David J. Poirier; Eric Boyd

    2003-01-20

    An oxygen-enriched furnace system for reduced CO2 and NOx emission has been developed. The furnace geometry, with a sidewall-mounted burner, was similar to configurations commonly encountered in a steel reheat furnace. The effect of stack oxygen concentration, oxygen enrichment level and air infiltration on fuel savings/CO2 reduction, NOx emissions and scale formation were investigated. The firing rate required to maintain the furnace temperature at 1100 C decreased linearly with increasing oxygen enrichment. At full oxygen enrichment a reduction of 40-45% in the firing rate was required to maintain furnace temperature. NOx emissions were relatively constant at oxygen enrichment levels below 60% and decreased concentration at all oxygen enrichment levels. Air infiltration also had an effect on NOx levels leading to emissions similar to those observed with no air infiltration but with similar stack oxygen concentrations. At high oxygen enrichment levels, there was a larger variation in the refractory surface-temperature on the roof and blind sidewall of the furnace. Scale habit, intactness, adhesion and oxidation rates were examined for five grades of steel over a range of stack oxygen concentrations and oxygen enrichment levels at 1100 degree C. The steel grade had the largest effect on scaling properties examined in this work. The stack oxygen concentration and the oxygen enrichment level had much smaller effects on the scaling properties.

  2. Oxygen enriched combustion system performance study. Phase 2: 100 percent oxygen enriched combustion in regenerative glass melters, Final report

    SciTech Connect (OSTI)

    Tuson, G.B.; Kobayashi, H.; Campbell, M.J.

    1994-08-01

    The field test project described in this report was conducted to evaluate the energy and environmental performance of 100% oxygen enriched combustion (100% OEC) in regenerative glass melters. Additional objectives were to determine other impacts of 100% OEC on melter operation and glass quality, and to verify on a commercial scale that an on-site Pressure Swing Adsorption oxygen plant can reliably supply oxygen for glass melting with low electrical power consumption. The tests constituted Phase 2 of a cooperative project between the United States Department of Energy, and Praxair, Inc. Phase 1 of the project involved market and technical feasibility assessments of oxygen enriched combustion for a range of high temperature industrial heating applications. An assessment of oxygen supply options for these applications was also performed during Phase 1, which included performance evaluation of a pilot scale 1 ton per day PSA oxygen plant. Two regenerative container glass melters were converted to 100% OEC operation and served as host sites for Phase 2. A 75 ton per day end-fired melter at Carr-Lowrey Glass Company in Baltimore, Maryland, was temporarily converted to 100% OEC in mid- 1990. A 350 tpd cross-fired melter at Gallo Glass Company in Modesto, California was rebuilt for permanent commercial operation with 100% OEC in mid-1991. Initially, both of these melters were supplied with oxygen from liquid storage. Subsequently, in late 1992, a Pressure Swing Adsorption oxygen plant was installed at Gallo to supply oxygen for 100% OEC glass melting. The particular PSA plant design used at Gallo achieves maximum efficiency by cycling the adsorbent beds between pressurized and evacuated states, and is therefore referred to as a Vacuum/Pressure Swing Adsorption (VPSA) plant.

  3. Uranium enrichment: investment options for the long term

    SciTech Connect (OSTI)

    Not Available

    1983-01-01

    The US government supplies a major portion of the enriched uranium used to fuel most of the nuclear power plants that furnish electricity in the free world. As manager of the US uranium enrichment concern, the Department of Energy (DOE) is investigating a number of technological choices to improve enrichment service and remain a significant world supplier. The Congress will ultimately select a strategy for federal investment in the uranium enrichment enterprise. A fundamental policy choice between possible future roles - that of the free world's main supplier of enrichment services, and that of a mainly domestic supplier - will underlie any investment decision the Congress makes. The technological choices are gaseous diffusion, gas centrifuge, and atomic vapor laser isotope separation (AVLIS). A base plan and four alternatives were examined by DOE and the Congressional Budget Office. In terms of total enterprise costs, Option IV, ultimately relying on advanced gas centrifuges for enrichment services, would offer the most economic approach, with costs over the full projection period totaling $123.5 billion. Option III, ultimately relying on AVLIS without gas centrifuge enrichment or gaseous diffusion, falls next in the sequence, with costs of $128.2 billion. Options I and II, involving combinations of the gas centrifuge and AVLIS technologies, follow closely with costs of $128.7 and $129.6 billion. The base plan has costs of $136.8 billion over the projection period. 1 figure, 22 tables.

  4. Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors

    SciTech Connect (OSTI)

    Not Available

    1993-07-01

    The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately.

  5. Report of Survey of Oak Ridge Isotope Enrichment (Calutron) Facility

    Office of Environmental Management (EM)

    Building 9204-3 | Department of Energy Report of Survey of Oak Ridge Isotope Enrichment (Calutron) Facility Building 9204-3 Report of Survey of Oak Ridge Isotope Enrichment (Calutron) Facility Building 9204-3 The purpose of this document is to report the results of a survey conducted at the Isotope Enrichment Facility (IEF, Calutron, Building 9204-3) on the Y-12 Plant property at the Oak Ridge Site. The survey was conducted during the week of November 29, 1999. The primary purpose of the

  6. Enriching stable isotopes: Alternative use for Urenco technology

    SciTech Connect (OSTI)

    Rakhorst, H.; de Jong, P.G.T.; Dawson, P.D.

    1996-12-31

    The International Urenco Group utilizes a technologically advanced centrifuge process to enrich uranium in the fissionable isotope {sup 235}U. The group operates plants in the United Kingdom, the Netherlands, and Germany and currently holds a 10% share of the multibillion dollar world enrichment market. In the early 1990s, Urenco embarked on a strategy of building on the company`s uniquely advanced centrifuge process and laser isotope separation (LIS) experience to enrich nonradioactive isotopes colloquially known as stable isotopes. This paper summarizes the present status of Urenco`s stable isotopes business.

  7. Development of an enrichment monitor for the Portsmouth GCEP

    SciTech Connect (OSTI)

    Strittmatter, R.B.; Stovall, L.A.; Sprinkle, J.K. Jr.

    1983-01-01

    We have developed a gas-phase UF/sub 6/ enrichment monitor for use by the International Atomic Energy Agency at the Portsmouth Gas Centrifuge Enrichment Plant. The enrichment monitoring system provides a method for effective nuclear materials accountability verification while reducing the effort for both the facility operator and the inspector. The experience with an inplant prototype monitor, the facility and operational constraints, and the constraints related to international safeguards inspection are described in terms of the impact on the monitor design.

  8. V

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Uzbekistan NNSA Partnership Successfully Removes All Remaining HEU from Uzbekistan WASHINGTON, DC - Today, the Department of Energy's National Nuclear Security Administration (DOE/NNSA) announced the successful return of the final 5 kilograms (approximately 11 pounds) of highly enriched uranium (HEU) spent fuel from the IIN-3M "Foton" research reactor in Tashkent, Uzbekistan to

    > 111 \il :^ a. \ii 1 1 < z a. 0 u. J < U iL 0 > h 0) or UJ > z D UCRL- 1 5 1 0

  9. Update on Fabrication of Extrusions for TREAT Trade Study

    SciTech Connect (OSTI)

    Luther, Erik Paul; Leckie, Rafael M.; Dombrowski, David E.; Papin, Pallas A.

    2014-03-05

    This supplemental report describes fuel fabrication efforts conducted for the Idaho National Laboratory Trade Study for the TREAT Conversion project that is exploring the replacement of the HEU (Highly Enriched Uranium) fuel core of the TREAT reactor with LEU (Low Enriched Uranium) fuel. Previous reports have documented fabrication of fuel by the upgrade process developed at Los Alamos National Laboratory. These experiments supplement an earlier report that describes efforts to increase the graphite content of extruded fuel and minimize cracking.

  10. Microsoft Word - DOE-ID-INL-16-032.docx

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    32 SECTION A. Project Title: European Mini-Plate Irradiation Experiment (EMPIRE) Capsules SECTION B. Project Description and Purpose: The European Mini-Plate Irradiation Experiment (EMPIRE) at Idaho National Laboratory (INL) includes fabrication and inspection, characterization, irradiation and post-irradiation examination (PIE) of monolithic and dispersion experimental fuel plates to support fuel development for the European reactor conversion from highly-enriched uranium (HEU) to low-enriched

  11. Higher Resolution Neutron Velocity Spectrometer Measurements of Enriched Uranium

    DOE R&D Accomplishments [OSTI]

    Rainwater, L. J.; Havens, W. W. Jr.

    1950-08-09

    The slow neutron transmission of a sample of enriched U containing 3.193 gm/cm2 was investigated with a resolution width of 1 microsec/m. Results of transmission measurements are shown graphically. (B.J.H.)

  12. Toxic Substances Control Act Uranium Enrichment Federal Facilities...

    Office of Environmental Management (EM)

    Thomas L. McCall, Jr. http:www.em.doe.govffaaortsca.html 4252001 Toxic Substances Control Act Uranium Enrichment Federal Facilities Compliance Agree.. Page 12 of 26 Deputy...

  13. The uranium cylinder assay system for enrichment plant safeguards

    SciTech Connect (OSTI)

    Miller, Karen A; Swinhoe, Martyn T; Marlow, Johnna B; Menlove, Howard O; Rael, Carlos D; Iwamoto, Tomonori; Tamura, Takayuki; Aiuchi, Syun

    2010-01-01

    Safeguarding sensitive fuel cycle technology such as uranium enrichment is a critical component in preventing the spread of nuclear weapons. A useful tool for the nuclear materials accountancy of such a plant would be an instrument that measured the uranium content of UF{sub 6} cylinders. The Uranium Cylinder Assay System (UCAS) was designed for Japan Nuclear Fuel Limited (JNFL) for use in the Rokkasho Enrichment Plant in Japan for this purpose. It uses total neutron counting to determine uranium mass in UF{sub 6} cylinders given a known enrichment. This paper describes the design of UCAS, which includes features to allow for unattended operation. It can be used on 30B and 48Y cylinders to measure depleted, natural, and enriched uranium. It can also be used to assess the amount of uranium in decommissioned equipment and waste containers. Experimental measurements have been carried out in the laboratory and these are in good agreement with the Monte Carlo modeling results.

  14. Ion laser isotope enrichment by photo-predissociation of formaldehyde

    DOE Patents [OSTI]

    Marling, John B.

    1977-06-17

    Enrichment of carbon, hydrogen and/or oxygen isotopes by means of isotopically selective photo-predissociation of formaldehyde is achieved by irradiation with a fixed frequency ion laser, specifically, a neon, cadmium, or xenon ion laser.

  15. Description of the Portsmouth Gas Centrifuge Enrichment Plant

    SciTech Connect (OSTI)

    Arthur, W.B.

    1980-12-16

    The Portsmouth Gas Centrifuge Enrichment Plant (GCEP) will be located at the site of the Portsmouth Gaseous Diffusion Plant in Piketon, Ohio. The purpose of the facility is to provide enriching services for the production of low assay enriched uranium for civilian nuclear power reactors. The construction and operation of the GCEP is administered by the US Department of Energy. The facility will be operated under contract from the US Government. Control of the GCEP rests solely with the US Government, which holds and controls access to the technology. Construction of GCEP is expected to be completed in the mid-1990's. Many facility design and operating procedures are subject to change. Nonetheless, the design described in this report does reflect current thinking. Descriptions of the general facility and major buildings such as the process buildings, feed and withdrawal building, cylinder storage and transfer, recycle/assembly building, and a summary of the centrifuge uranium enriching process are provided in this report.

  16. Evaluating the necessity of certain gas centrifuge enrichment plant process

    Office of Scientific and Technical Information (OSTI)

    parameters to a safeguards approach that includes possible remote inspections driven by information (Conference) | SciTech Connect Evaluating the necessity of certain gas centrifuge enrichment plant process parameters to a safeguards approach that includes possible remote inspections driven by information Citation Details In-Document Search Title: Evaluating the necessity of certain gas centrifuge enrichment plant process parameters to a safeguards approach that includes possible remote

  17. Evaluating the necessity of certain gas centrifuge enrichment plant process

    Office of Scientific and Technical Information (OSTI)

    parameters to a safeguards approach that includes possible remote inspections driven by information (Conference) | SciTech Connect Evaluating the necessity of certain gas centrifuge enrichment plant process parameters to a safeguards approach that includes possible remote inspections driven by information Citation Details In-Document Search Title: Evaluating the necessity of certain gas centrifuge enrichment plant process parameters to a safeguards approach that includes possible remote

  18. Evaluating the necessity of certain gas centrifuge enrichment plant process

    Office of Scientific and Technical Information (OSTI)

    parameters to a safeguards approach that includes possible remote inspections driven by information (Conference) | SciTech Connect Evaluating the necessity of certain gas centrifuge enrichment plant process parameters to a safeguards approach that includes possible remote inspections driven by information Citation Details In-Document Search Title: Evaluating the necessity of certain gas centrifuge enrichment plant process parameters to a safeguards approach that includes possible remote

  19. Evaluating the necessity of certain gas centrifuge enrichment plant process

    Office of Scientific and Technical Information (OSTI)

    parameters to a safeguards approach that includes possible remote inspections driven by information (Conference) | SciTech Connect Evaluating the necessity of certain gas centrifuge enrichment plant process parameters to a safeguards approach that includes possible remote inspections driven by information Citation Details In-Document Search Title: Evaluating the necessity of certain gas centrifuge enrichment plant process parameters to a safeguards approach that includes possible remote

  20. News Media invited to interview Jefferson Lab summer science enrichment

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    program participants; cover closing Poster Session | Jefferson Lab News Media invited to interview Jefferson Lab summer science enrichment program participants; cover closing Poster Session August 1, 2007 News Media representatives are invited to interview, photograph and/or film participants of Jefferson Lab's summer science enrichment programs as they share their summer experiences and projects with JLab staff during a Poster Session scheduled for Friday, Aug. 3, from 11:30 a.m.-1:30 p.m.

  1. Jefferson Lab Seeks Applicants for Science Teacher Enrichment Program |

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Jefferson Lab Seeks Applicants for Science Teacher Enrichment Program Jefferson Lab Seeks Applicants for Science Teacher Enrichment Program March 14, 2001 Calling all middle school teachers who instruct science classes. Jefferson Lab would like to help you refresh and hone your science knowledge and teaching skills over the summer. The Department of Energy physics research lab, located at 12000 Jefferson Ave. in Newport News, is seeking applications for its four-week, summer physics

  2. Jefferson Lab seeks applicants for summer, science teacher enrichment

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    program | Jefferson Lab seeks applicants for summer, science teacher enrichment program Jefferson Lab seeks applicants for summer, science teacher enrichment program February 26, 2003 Calling all middle school teachers who instruct science classes. Jefferson Lab would like to help you refresh and hone your science knowledge and teaching skills over the summer. The Department of Energy physics research lab, located at 12000 Jefferson Ave. in Newport News, is seeking applications for its

  3. Laboratory Enrichment of Radioactive Assemblages and Estimation of Thorium

    Office of Scientific and Technical Information (OSTI)

    and Uranium Radioactivity in Fractions Separated from Placer Sands in Southeast Bangladesh (Journal Article) | SciTech Connect SciTech Connect Search Results Journal Article: Laboratory Enrichment of Radioactive Assemblages and Estimation of Thorium and Uranium Radioactivity in Fractions Separated from Placer Sands in Southeast Bangladesh Citation Details In-Document Search Title: Laboratory Enrichment of Radioactive Assemblages and Estimation of Thorium and Uranium Radioactivity in

  4. Energy Department Selects Global Laser Enrichment for Future Operations at

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Paducah Site | Department of Energy Global Laser Enrichment for Future Operations at Paducah Site Energy Department Selects Global Laser Enrichment for Future Operations at Paducah Site November 27, 2013 - 12:00pm Addthis Workers inspect cylinders containing depleted uranium hexafluoride. Workers inspect cylinders containing depleted uranium hexafluoride. Media Contact (202) 586-4940 Washington, D.C. - The U.S. Department of Energy announced today that it will open negotiations with Global

  5. College Bound American Indian Math and Science Enrichment Program (AIMS). Final report

    SciTech Connect (OSTI)

    1998-09-01

    Southwestern Indian Polytechnic Institute (SIPI), was founded in 1971 and is located on 164 acres in northwest Albuquerque, New Mexico in the center of New Mexico`s agricultural and high-tech corridors. SIPI became accredited as a community college in 1993, serves Native Americans nationwide, and is governed by a nationally-tribally appointed Board of Regents (Jicarilla Apache, Joint Oklahoma Tribes, Mescalero Apache, Navajo Nation-Arizona, Navajo Nation-New Mexico, Ten Southern Pueblos, and Eight Northern Pueblos, Southern Ute, Inter-tribal Council of Arizona, and Oglala Sioux). In 1993, The US Department of Education, TRIO Programs no longer funded the Southwestern Indian Polytechnic Institute (SIPI) Summer Math and Science Enrichment Program. However, with US Department of Energy funding SIPI was able to continue service to the Native American community under the new title of College Bound American Indian Math and Science (AIMS) Enrichment Program. This new program continued the goals and objectives of the TRIO program with an expanded focus that included students from more Native American communities nationwide. The program also interfaced with a teacher enrichment program (Rural American Indian Science Education-RAISE) sponsored by the Bureau of Indian Affairs and Sandia National Labs (SNL). SIPI in collaboration with Sandia National Laboratories and Lawrence Livermore National (LLNL) Laboratory established a mathematics and science enrichment program at SIPI for students attending rural high schools serving predominantly Native American populations. The primary goal of the program was to provide 9th--12th grade students, mostly Native American, the skills and knowledge, interest and motivation, and strategies to remain in high school and pursue a college education in a math, science, or technology based field. Each year, the program included a six-week intensive residential summer program located at SIPI as well as academic year support activities at the student`s high school. A summary of the program activities during the grant period is given.

  6. Jefferson Lab Offers Science Enrichment Program for 5th, 6th...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Fifty-four teachers who participated in enrichment programs at JLab for teachers of science presented the activities and demonstrations. Jefferson Lab Offers Science Enrichment ...

  7. Jefferson Lab Offers Science Enrichment Program for 5th, 6th...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Jefferson Lab Offers Science Enrichment Program for 5th, 6th & 8th Grade Teachers NEWPORT ... accepting applications for its science enrichment program for fifth-, sixth- and ...

  8. Influence of hexavalent chromium on lactate-enriched Hanford groundwater microbial communities.

    SciTech Connect (OSTI)

    Somenahally, Anil C; Mosher, Jennifer J; Yuan, Tong; Podar, Mircea; Phelps, Tommy Joe; Brown, Steven D; Yang, Zamin Koo; Hazen, Terry C; Arkin, Adam; Palumbo, Anthony Vito; Zhou, Jizhong; Elias, Dwayne A

    2013-01-01

    Microbial reduction and immobilization of chromate (Cr(VI)) is a plausible bioremediation strategy. However, higher Cr(VI) concentrations may impose stress on native Cr-reducing communities. We sought to determine if Cr(VI) would influence the lactate enriched native microbial community structure and function in groundwater from the Cr contaminated site at Hanford, WA. Steady state continuous flow bioreactors were amended with lactate and Cr(VI) (0.0, 0.1 and 3.0 mg/L). Microbial growth, metabolites, Cr(VI) concentrations, 16S rRNA gene sequences and GeoChip based functional gene composition in bioreactors were monitored for 15 weeks. Temporal trends and some differences in growth, metabolite profiles, and community composition were observed, largely between Low-Cr and High-Cr bioreactors. In both High-Cr and Low-Cr bioreactors, Cr(VI) was reduced in the bioreactors. With lactate enrichment, the native communities did not significantly differ between Cr concentrations. Native bacterial communities were diverse, whereas after lactate enrichment, Pelosinus spp., and Sporotalea spp., were the most predominant groups in all bioreactors. Similarly, the Archaea diversity significantly decreased from Methanosaeta (35%), Methanosarcina (17%), Halobacteriales (12%), Methanoregula (8%) and others, to mostly Methanosarcina spp. (95%) after lactate enrichment. Composition of several key functional genes was distinct in Low-Cr bioreactors compared to High-Cr. Among the Cr resistant probes (chrA), Burkholderia vietnamiensis, Comamonas testosterone and Ralstonia pickettii proliferated in Cr amended bioreactors. In-situ fermentative conditions facilitated Cr(VI) reduction, and as a result the 3.0 mg/L Cr(VI) did not appear to give chromate reducing strains a competitive advantage for proliferation or for increasing Cr-reduction.

  9. Prussian Blue Nanoparticles for the Enrichment of Radioactive Cesium in Solutions - 13275

    SciTech Connect (OSTI)

    Parajuli, Durga; Kitajima, Akiko; Tanaka, Hisashi; Kawamoto, Tohru

    2013-07-01

    Prussian blue (PB) nanoparticles in different form were studied for the adsorptive enrichment of Cs in solutions. Water dispersible nano-PB was found to be highly effective on removing trace level Cs in stagnant waters. The nano-PB loaded filters were effective on collecting Cs in flow systems like river water, thus provides a big relief on controlling the environmental mobility of Cs and its entry to the productive lands via water. Water insoluble nano-PB adsorbent possesses high Cs loading capacity and selectivity and it is found to be the ultimate option for the systems containing high concentration Cs. (authors)

  10. Generalized Modeling of Enrichment Cascades That Include Minor Isotopes

    SciTech Connect (OSTI)

    Weber, Charles F

    2012-01-01

    The monitoring of enrichment operations may require innovative analysis to allow for imperfect or missing data. The presence of minor isotopes may help or hurt - they can complicate a calculation or provide additional data to corroborate a calculation. However, they must be considered in a rigorous analysis, especially in cases involving reuse. This study considers matched-abundanceratio cascades that involve at least three isotopes and allows generalized input that does not require all feed assays or the enrichment factor to be specified. Calculations are based on the equations developed for the MSTAR code but are generalized to allow input of various combinations of assays, flows, and other cascade properties. Traditional cascade models have required specification of the enrichment factor, all feed assays, and the product and waste assays of the primary enriched component. The calculation would then produce the numbers of stages in the enriching and stripping sections and the remaining assays in waste and product streams. In cases where the enrichment factor or feed assays were not known, analysis was difficult or impossible. However, if other quantities are known (e.g., additional assays in waste or product streams), a reliable calculation is still possible with the new code, but such nonstandard input may introduce additional numerical difficulties into the calculation. Thus, the minimum input requirements for a stable solution are discussed, and a sample problem with a non-unique solution is described. Both heuristic and mathematically required guidelines are given to assist the application of cascade modeling to situations involving such non-standard input. As a result, this work provides both a calculational tool and specific guidance for evaluation of enrichment cascades in which traditional input data are either flawed or unknown. It is useful for cases involving minor isotopes, especially if the minor isotope assays are desired (or required) to be important contributors to the overall analysis.

  11. Conversion of Worcester Polytechnic Institute Reactor to low enriched uranium (LEU) fuel: Technical progress report for period August 15, 1987-February 15, 1988

    SciTech Connect (OSTI)

    Newton, T.H. Jr.

    1988-02-01

    An HEU fuel element was removed from the WPI core and tested in a Babcock-Wilcox 6M shipping container on August 27, 1987, for radiation level adequacy in shipping. Levels were found to be adequate so that use of the 6M container can be made in shipping the HEU fuel after a few weeks of decay time. A final submittal of the SAR technical specification changes relating to the fuel conversion was made on September 17, 1987. Questions regarding this submittal were received on January 25, 1988, and responses to these questions were made on February 10, 1988.

  12. Microbial gene functions enriched in the Deepwater Horizon deep-sea oil plume

    SciTech Connect (OSTI)

    Lu, Z.; Deng, Y.; Nostrand, J.D. Van; He, Z.; Voordeckers, J.; Zhou, A.; Lee, Y.-J.; Mason, O.U.; Dubinsky, E.; Chavarria, K.; Tom, L.; Fortney, J.; Lamendella, R.; Jansson, J.K.; D?haeseleer, P.; Hazen, T.C.; Zhou, J.

    2011-06-15

    The Deepwater Horizon oil spill in the Gulf of Mexico is the deepest and largest offshore spill in U.S. history and its impacts on marine ecosystems are largely unknown. Here, we showed that the microbial community functional composition and structure were dramatically altered in a deep-sea oil plume resulting from the spill. A variety of metabolic genes involved in both aerobic and anaerobic hydrocarbon degradation were highly enriched in the plume compared to outside the plume, indicating a great potential for intrinsic bioremediation or natural attenuation in the deep-sea. Various other microbial functional genes relevant to carbon, nitrogen, phosphorus, sulfur and iron cycling, metal resistance, and bacteriophage replication were also enriched in the plume. Together, these results suggest that the indigenous marine microbial communities could play a significant role in biodegradation of oil spills in deep-sea environments.

  13. EXTERNAL PHOTOEVAPORATION OF THE SOLAR NEBULA: JUPITER's NOBLE GAS ENRICHMENTS

    SciTech Connect (OSTI)

    Monga, Nikhil; Desch, Steven

    2015-01-01

    We present a model explaining the elemental enrichments in Jupiter's atmosphere, particularly the noble gases Ar, Kr, and Xe. While He, Ne, and O are depleted, seven other elements show similar enrichments (∼3 times solar, relative to H). Being volatile, Ar is difficult to fractionate from H{sub 2}. We argue that external photoevaporation by far-ultraviolet (FUV) radiation from nearby massive stars removed H{sub 2}, He, and Ne from the solar nebula, but Ar and other species were retained because photoevaporation occurred at large heliocentric distances where temperatures were cold enough (≲ 30 K) to trap them in amorphous water ice. As the solar nebula lost H, it became relatively and uniformly enriched in other species. Our model improves on the similar model of Guillot and Hueso. We recognize that cold temperatures alone do not trap volatiles; continuous water vapor production is also necessary. We demonstrate that FUV fluxes that photoevaporated the disk generated sufficient water vapor in regions ≲ 30 K to trap gas-phase species in amorphous water ice in solar proportions. We find more efficient chemical fractionation in the outer disk: whereas the model of Guillot and Hueso predicts a factor of three enrichment when only <2% of the disk mass remains, we find the same enrichments when 30% of the disk mass remains. Finally, we predict the presence of ∼0.1 M {sub ⊕} of water vapor in the outer solar nebula and protoplanetary disks in H II regions.

  14. Signatures and Methods for the Automated Nondestructive Assay of UF6 Cylinders at Uranium Enrichment Plants

    SciTech Connect (OSTI)

    Smith, Leon E.; Mace, Emily K.; Misner, Alex C.; Shaver, Mark W.

    2010-08-08

    International Atomic Energy Agency (IAEA) inspectors currently perform periodic inspections at uranium enrichment plants to verify UF6 cylinder enrichment declarations. Measurements are typically performed with handheld high-resolution sensors on a sampling of cylinders taken to be representative of the facility’s entire cylinder inventory. These measurements are time-consuming, expensive, and assay only a small fraction of the total cylinder volume. An automated nondestructive assay system capable of providing enrichment measurements over the full volume of the cylinder could improve upon current verification practices in terms of manpower and assay accuracy. Such a station would use sensors that can be operated in an unattended mode at an industrial facility: medium-resolution scintillators for gamma-ray spectroscopy (e.g., NaI(Tl)) and moderated He-3 neutron detectors. This sensor combination allows the exploitation of additional, more-penetrating signatures beyond the traditional 185-keV emission from U-235: neutrons produced from F-19(α,n) reactions (spawned primarily from U 234 alpha emission) and high-energy gamma rays (extending up to 8 MeV) induced by neutrons interacting in the steel cylinder. This paper describes a study of these non-traditional signatures for the purposes of cylinder enrichment verification. The signatures and the radiation sensors designed to collect them are described, as are proof-of-principle cylinder measurements and analyses. Key sources of systematic uncertainty in the non-traditional signatures are discussed, and the potential benefits of utilizing these non-traditional signatures, in concert with an automated form of the traditional 185-keV-based assay, are discussed.

  15. PERSPECTIVES ON INTRACLUSTER ENRICHMENT AND THE STELLAR INITIAL MASS FUNCTION IN ELLIPTICAL GALAXIES

    SciTech Connect (OSTI)

    Loewenstein, Michael

    2013-08-10

    Stars formed in galaxy cluster potential wells must be responsible for the high level of enrichment measured in the intracluster medium (ICM); however, there is increasing tension between this truism and the parsimonious assumption that the stars in the generally old population studied optically in cluster galaxies emerged from the same formation sites at the same epochs. We construct a phenomenological cluster enrichment model to demonstrate that ICM elemental abundances are underestimated by a factor >2 for standard assumptions about the stellar population-a discrepancy we call the ''cluster elemental abundance paradox''. Recent evidence of an elliptical galaxy initial mass function (IMF) skewed to low masses deepens the paradox. We quantify the adjustments to the star formation efficiency and IMF, and Type Ia supernovae (SNIa) production efficiency, required to resolve this while being consistent with the observed ICM abundance pattern. The necessary enhancement in metal enrichment may, in principle, originate in the observed stellar population if a larger fraction of stars in the supernova-progenitor mass range form from an IMF that is either bottom-light or top-heavy, with the latter in some conflict with observed ICM abundance ratios. Other alternatives that imply more modest revisions to the IMF, mass return and remnant fractions, and primordial fraction, posit an increase in the fraction of 3-8 M{sub Sun} stars that explode as SNIa or assume that there are more stars than conventionally thought-although the latter implies a high star formation efficiency. We discuss the feasibility of these various solutions and the implications for the diversity of star formation in the universe, the process of elliptical galaxy formation, and the origin of this ''hidden'' source of ICM metal enrichment.

  16. Commissioning Measurements and Experience Obtained from the Installation of a Fissile Mass Flow monitor in the URAL Electrochemical Integrated Plant (UEIP) in Novouralsk

    SciTech Connect (OSTI)

    March-Leuba, J.; Mastal, E.; Powell, D.; Sumner, J.; Uckan, T.; Vines, V.

    1999-07-25

    The Blend Down Monitoring System (BDMS) equipment sent earlier to the Ural Electrochemical Integrated Plant (UEIP) at Novouralsk, Russia, was installed and implemented successfully on February 2, 1999. The BDMS installation supports the highly enriched uranium (HEU) Transparency Implementation Program for material subject to monitoring under the HEU purchase agreement between the United States of America (USA) and the Russian Federation (RF). The BDMS consists of the Oak Ridge National Laboratory (ORNL) Fissile (uranium-235) Mass Flow Monitor (FMFM) and the Los Alamos National Laboratory (LANL) Enrichment Monitor (EM). Two BDMS?s for monitoring the Main and Reserve HEU blending process lines were installed at UEIP. Independent operation of the FMFM Main and FMFM Reserve was successfully demonstrated for monitoring the fissile mass flow as well as the traceability of HEU to the product low enriched uranium. The FMFM systems failed when both systems were activated during the calibration phase due to a synchronization problem between the systems. This operational failure was caused by the presence of strong electromagnetic interference (EMI) in the blend point. The source-modulator shutter motion of the two FMFM systems was not being properly synchronized because of EMI producing a spurious signal on the synchronization cable connecting the two FMFM cabinets. The signature of this failure was successfully reproduced at ORNL after the visit. This unexpected problem was eliminated by a hardware modification and software improvements during a recent visit (June 9-11, 1999) to UEIP, and both systems are now operating as expected.

  17. Perimeter safeguards techniques for uranium-enrichment plants

    SciTech Connect (OSTI)

    Fehlau, P.E.; Chamber, W.H.

    1981-09-01

    In 1972, a working group of the International Atomic Energy Agency identified a goal to develop and evaluate perimeter safeguards for uranium isotope enrichment plants. As part of the United State's response to that goal, Los Alamos Detection and Verification personnel studied gamma-ray and neutron emissions from uranium hexafluoride. They developed instruments that use the emissions to verify uranium enrichment and to monitor perimeter personnel and shipping portals. Unattended perimeter monitors and hand-held verification instruments were evaluated in field measurements and, when possible, were loaned to enrichment facilities for trials. None of the seven package monitoring techniques that were investigated proved entirely satisfactory for an unattended monitor. They either revealed proprietary information about centrifuge design or were subject to interference by shielding materials that could be present in a package. Further evaluation in a centrifuge facility may help in developing an acceptable attended package monitor. 34 figures, 9 tables.

  18. Centrifuge enrichment plants. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect (OSTI)

    Not Available

    1994-03-01

    The bibliography contains citations concerning the design, control, monitoring, and safety of centrifuge enrichment plants. Power supplies, enrichment plant safeguards, facility design, cascade heater test loops to monitor the enrichment process, inspection strategies, and the socioeconomic effects of centrifuge enrichment plants are examined. Radioactive waste disposal problems are considered. (Contains a minimum of 172 citations and includes a subject term index and title list.)

  19. Centrifuge enrichment plants. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect (OSTI)

    1993-09-01

    The bibliography contains citations concerning the design, control, monitoring, and safety of centrifuge enrichment plants. Power supplies, enrichment plant safeguards, facility design, cascade heater test loops to monitor the enrichment process, inspection strategies, and the socioeconomic effects of centrifuge enrichment plants are examined. Radioactive waste disposal problems are considered. (Contains a minimum of 171 citations and includes a subject term index and title list.)

  20. Sonication-based isolation and enrichment of Chlorella protothecoides chloroplasts for illumina genome sequencing

    SciTech Connect (OSTI)

    Angelova, Angelina; Park, Sang-Hycuk; Kyndt, John; Fitzsimmons, Kevin; Brown, Judith K

    2013-09-01

    With the increasing world demand for biofuel, a number of oleaginous algal species are being considered as renewable sources of oil. Chlorella protothecoides Krüger synthesizes triacylglycerols (TAGs) as storage compounds that can be converted into renewable fuel utilizing an anabolic pathway that is poorly understood. The paucity of algal chloroplast genome sequences has been an important constraint to chloroplast transformation and for studying gene expression in TAGs pathways. In this study, the intact chloroplasts were released from algal cells using sonication followed by sucrose gradient centrifugation, resulting in a 2.36-fold enrichment of chloroplasts from C. protothecoides, based on qPCR analysis. The C. protothecoides chloroplast genome (cpDNA) was determined using the Illumina HiSeq 2000 sequencing platform and found to be 84,576 Kb in size (8.57 Kb) in size, with a GC content of 30.8 %. This is the first report of an optimized protocol that uses a sonication step, followed by sucrose gradient centrifugation, to release and enrich intact chloroplasts from a microalga (C. prototheocoides) of sufficient quality to permit chloroplast genome sequencing with high coverage, while minimizing nuclear genome contamination. The approach is expected to guide chloroplast isolation from other oleaginous algal species for a variety of uses that benefit from enrichment of chloroplasts, ranging from biochemical analysis to genomics studies.

  1. Safeguards Verification Measurements using Laser Ablation, Absorbance Ratio Spectrometry in Gaseous Centrifuge Enrichment Plants

    SciTech Connect (OSTI)

    Anheier, Norman C.; Cannon, Bret D.; Kulkarni, Gourihar R.; Munley, John T.; Nelson, Danny A.; Qiao, Hong; Phillips, Jon R.

    2012-07-17

    Laser Ablation Absorbance Ratio Spectrometry (LAARS) is a new verification measurement technology under development at the US Department of Energy (DOE) Pacific Northwest National Laboratory (PNNL). LAARS uses three lasers to ablate and then measure the relative isotopic abundance of uranium compounds. An ablation laser is tightly focused on uranium-bearing solids, producing a small atomic uranium vapor plume. Two collinear wavelength-tuned spectrometry lasers transit through the plume and the absorbance of U-235 and U-238 isotopes are measured to determine U-235 enrichment. The measurement is independent of chemical form and degree of dilution with nuisance dust and other materials. LAARS has high relative precision and detection limits approaching the femtogram range for U-235. The sample is scanned and assayed point-by-point at rates reaching 1 million measurements/hour, enabling LAARS to detect and analyze uranium in trace samples. The spectrometer is assembled using primarily commercially available components and features a compact design and automated analysis.Two specific gaseous centrifuge enrichment plant (GCEP) applications of the spectrometer are currently under development: 1) LAARS-Environmental Sampling (ES), which collects and analyzes aerosol particles for GCEP misuse detection and 2) LAARS-Destructive Assay (DA), which enables onsite enrichment DA sample collection and analysis for protracted diversion detection. The two applications propose game-changing technological advances in GCEP safeguards verification.

  2. Blend Down Monitoring System Fissile Mass Flow Monitor Implementation at the ElectroChemical Plant, Zelenogorsk, Russia

    SciTech Connect (OSTI)

    Uckan, T.

    2005-11-11

    The implementation plans and preparations for installation of the Fissile Mass Flow Monitor (FMFM) equipment at the ElectroChemical Plant (ECP), Zelenogorsk, Russia, are presented in this report. The FMFM, developed at Oak Ridge National Laboratory, is part of the Blend Down Monitoring System (BDMS), developed for the U.S. Department of Energy Highly Enriched Uranium (HEU) Transparency Implementation Program. The BDMS provides confidence to the United States that the Russian nuclear facilities supplying the lower-assay ({approx}4%) product low enriched uranium (P-LEU) to the United States from down-blended weapons-grade HEU are meeting the nonproliferation goals of the government-to-government HEU Purchase Agreement, signed between the Russian Federation and the United States in 1993. The first BDMS has been operational at Ural Electrochemical Integrated Plant, Novouralsk, since February 1999 and is successfully providing HEU transparency data to the United States. The second BDMS was installed at ECP in February 2003. The FMFM makes use of a set of thermalized californium-252 ({sup 252}Cf) spontaneous neutron sources for a modulated fission activation of the UF{sub 6} gas stream for measuring the {sup 235}U fissile mass flow rate. To do this, the FMFM measures the transport time of the fission fragments created from the fission activation process under the modulated source to the downstream detectors by detecting the delayed gamma rays from the fission fragments. The FMFM provides unattended, nonintrusive measurements of the {sup 235}U mass flow in the HEU, LEU blend stock, and P-LEU process legs. The FMFM also provides the traceability of the HEU flow to the product process leg. This report documents the technical installation requirements and the expected operational characteristics of the ECP FMFM.

  3. INSIGHTS INTO PRE-ENRICHMENT OF STAR CLUSTERS AND SELF-ENRICHMENT OF DWARF GALAXIES FROM THEIR INTRINSIC METALLICITY DISPERSIONS

    SciTech Connect (OSTI)

    Leaman, Ryan

    2012-12-01

    Star clusters are known to have smaller intrinsic metallicity spreads than dwarf galaxies due to their shorter star formation timescales. Here we use individual spectroscopic [Fe/H] measurements of stars in 19 Local Group dwarf galaxies, 13 Galactic open clusters, and 49 globular clusters to show that star cluster and dwarf galaxy linear metallicity distributions are binomial in form, with all objects showing strong correlations between their mean linear metallicity Z-bar and intrinsic spread in metallicity {sigma}(Z){sup 2}. A plot of {sigma}(Z){sup 2} versus Z-bar shows that the correlated relationships are offset for the dwarf galaxies from the star clusters. The common binomial nature of these linear metallicity distributions can be explained with a simple inhomogeneous chemical evolution model, where the star cluster and dwarf galaxy behavior in the {sigma}(Z){sup 2}- Z-bar diagram is reproduced in terms of the number of enrichment events, covering fraction, and intrinsic size of the enriched regions. The inhomogeneity of the self-enrichment sets the slope for the observed dwarf galaxy {sigma}(Z){sup 2}- Z-bar correlation. The offset of the star cluster sequence from that of the dwarf galaxies is due to pre-enrichment, and the slope of the star cluster sequence represents the remnant signature of the self-enriched history of their host galaxies. The offset can be used to separate star clusters from dwarf galaxies without a priori knowledge of their luminosity or dynamical mass. The application of the inhomogeneous model to the {sigma}(Z){sup 2}- Z-bar relationship provides a numerical formalism to connect the self-enrichment and pre-enrichment between star clusters and dwarf galaxies using physically motivated chemical enrichment parameters. Therefore we suggest that the {sigma}(Z){sup 2}- Z-bar relationship can provide insight into what drives the efficiency of star formation and chemical evolution in galaxies, and is an important prediction for galaxy simulation models to reproduce.

  4. Four-Year Plan | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    HEU Removal South Africa HEU Removal Sweden Plutonium Removal Taiwan HEU Removal Turkey HEU Removal Ukraine HEU Removal United Kingdom HEU Removal Uzbekistan HEU Removal...

  5. CRAD, Management- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Management program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  6. Paducah Plant Begins Enrichment Operations after Five Parties Strike Agreement

    Broader source: Energy.gov [DOE]

    PADUCAH, Ky. – On this Friday, June 1, the first of about 1,100 14-ton cylinders filled with depleted uranium tails will be re-fed into the Paducah Gaseous Diffusion Plant, officially marking the start of a project that will extend enrichment operations there for a year.

  7. CRAD, Training- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Training Program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  8. Microchip method for the enrichment of specific DNA sequences

    DOE Patents [OSTI]

    Mirzabekov, A.D.; Lysov, Y.P.; Shick, V.V.; Dubiley, S.A.

    1998-12-22

    A method for enriching specific genetic material sequences is provided, whereby oligonucleotide molecules complementary to the desired genetic material is first used to isolate the genetic material from a first source of genomic material. Then the genetic material is used as a label to isolate similar genetic sequences from other sources. 4 figs.

  9. Neutron spectrometry for UF6 enrichment verification in storage cylinders

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Mengesha, Wondwosen; Kiff, Scott D.

    2015-01-29

    Verification of declared UF6 enrichment and mass in storage cylinders is of great interest in nuclear material nonproliferation. Nondestructive assay (NDA) techniques are commonly used for safeguards inspections to ensure accountancy of declared nuclear materials. Common NDA techniques used include gamma-ray spectrometry and both passive and active neutron measurements. In the present study, neutron spectrometry was investigated for verification of UF6 enrichment in 30B storage cylinders based on an unattended and passive measurement approach. MCNP5 and Geant4 simulated neutron spectra, for selected UF6 enrichments and filling profiles, were used in the investigation. The simulated neutron spectra were analyzed using principalmore » component analysis (PCA). The PCA technique is a well-established technique and has a wide area of application including feature analysis, outlier detection, and gamma-ray spectral analysis. Results obtained demonstrate that neutron spectrometry supported by spectral feature analysis has potential for assaying UF6 enrichment in storage cylinders. The results from the present study also showed that difficulties associated with the UF6 filling profile and observed in other unattended passive neutron measurements can possibly be overcome using the approach presented.« less

  10. Genetic engineering of syringyl-enriched lignin in plants

    DOE Patents [OSTI]

    Chiang, Vincent Lee; Li, Laigeng

    2004-11-02

    The present invention relates to a novel DNA sequence, which encodes a previously unidentified lignin biosynthetic pathway enzyme, sinapyl alcohol dehydrogenase (SAD) that regulates the biosynthesis of syringyl lignin in plants. Also provided are methods for incorporating this novel SAD gene sequence or substantially similar sequences into a plant genome for genetic engineering of syringyl-enriched lignin in plants.

  11. Microchip method for the enrichment of specific DNA sequences

    DOE Patents [OSTI]

    Mirzabekov, Andrei Darievich; Lysov, Yuri Petrovich; Shick, Valentine Vladimirovich; Dubiley, Svetlana Alekseevna

    1998-01-01

    A method for enriching specific genetic material sequences is provided, whereby oligonucleotide molecules complementary to the desired genetic material is first used to isolate the genetic material from a first source of genomic material. Then the genetic material is used as a label to isolate similar genetic sequences from other sources.

  12. Laser Isotope Enrichment for Medical and Industrial Applications

    SciTech Connect (OSTI)

    Leonard Bond

    2006-07-01

    Laser Isotope Enrichment for Medical and Industrial Applications by Jeff Eerkens (University of Missouri), Jay Kunze (Idaho State University), and Leonard Bond (Idaho National Laboratory) The principal isotope enrichment business in the world is the enrichment of uranium for commercial power reactor fuels. However, there are a number of other needs for separated isotopes. Some examples are: 1) Pure isotopic targets for irradiation to produce medical radioisotopes. 2) Pure isotopes for semiconductors. 3) Low neutron capture isotopes for various uses in nuclear reactors. 4) Isotopes for industrial tracer/identification applications. Examples of interest to medicine are targets to produce radio-isotopes such as S-33, Mo-98, Mo-100, W-186, Sn-112; while for MRI diagnostics, the non-radioactive Xe-129 isotope is wanted. For super-semiconductor applications some desired industrial isotopes are Si-28, Ga-69, Ge-74, Se-80, Te-128, etc. An example of a low cross section isotope for use in reactors is Zn-68 as a corrosion inhibitor material in nuclear reactor primary systems. Neutron activation of Ar isotopes is of interest in industrial tracer and diagnostic applications (e.g. oil-logging). . In the past few years there has been a sufficient supply of isotopes in common demand, because of huge Russian stockpiles produced with old electromagnetic and centrifuge separators previously used for uranium enrichment. Production of specialized isotopes in the USA has been largely accomplished using old ”calutrons” (electromagnetic separators) at Oak Ridge National Laboratory. These methods of separating isotopes are rather energy inefficient. Use of lasers for isotope separation has been considered for many decades. None of the proposed methods have attained sufficient proof of principal status to be economically attractive to pursue commercially. Some of the authors have succeeded in separating sulfur isotopes using a rather new and different method, known as condensation repression. In this scheme a gas, of the selected isotopes for enrichment, is irradiated with a laser at a particular wavelength that would excite only one of the isotopes. The entire gas is subject to low temperatures sufficient to cause condensation on a cold surface. Those molecules in the gas that the laser excited are not as likely to condense as are the unexcited molecules. Hence the gas drawn out of the system will be enriched in the isotope that was excited by the laser. We have evaluated the relative energy required in this process if applied on a commercial scale. We estimate the energy required for laser isotope enrichment is about 20% of that required in centrifuge separations, and 2% of that required by use of "calutrons".

  13. Approach to proliferation risk assessment based on multiple objective analysis framework

    SciTech Connect (OSTI)

    Andrianov, A.; Kuptsov, I.

    2013-07-01

    The approach to the assessment of proliferation risk using the methods of multi-criteria decision making and multi-objective optimization is presented. The approach allows the taking into account of the specifics features of the national nuclear infrastructure, and possible proliferation strategies (motivations, intentions, and capabilities). 3 examples of applying the approach are shown. First, the approach has been used to evaluate the attractiveness of HEU (high enriched uranium)production scenarios at a clandestine enrichment facility using centrifuge enrichment technology. Secondly, the approach has been applied to assess the attractiveness of scenarios for undeclared production of plutonium or HEU by theft of materials circulating in nuclear fuel cycle facilities and thermal reactors. Thirdly, the approach has been used to perform a comparative analysis of the structures of developing nuclear power systems based on different types of nuclear fuel cycles, the analysis being based on indicators of proliferation risk.

  14. Update On The Development, Testing, And Manufacture Of High Density LEU-Foil Targets For The Production Of Mo-99

    SciTech Connect (OSTI)

    Creasy, John T

    2015-05-12

    This project has the objective to reduce and/or eliminate the use of HEU in commerce. Steps in the process include developing a target testing methodology that is bounding for all Mo-99 target irradiators, establishing a maximum target LEU-foil mass, developing a LEU-foil target qualification document, developing a bounding target failure analysis methodology (failure in reactor containment), optimizing safety vs. economics (goal is to manufacture a safe, but relatively inexpensive target to offset the inherent economic disadvantage of using LEU in place of HEU), and developing target material specifications and manufacturing QC test criteria. The slide presentation is organized under the following topics: Objective, Process Overview, Background, Team Structure, Key Achievements, Experiment and Activity Descriptions, and Conclusions. The High Density Target project has demonstrated: approx. 50 targets irradiated through domestic and international partners; proof of concept for two front end processing methods; fabrication of uranium foils for target manufacture; quality control procedures and steps for manufacture; multiple target assembly techniques; multiple target disassembly devices; welding of targets; thermal, hydraulic, and mechanical modeling; robust target assembly parametric studies; and target qualification analysis for insertion into very high flux environment. The High Density Target project has tested and proven several technologies that will benefit current and future Mo-99 producers.

  15. Identification of c-Type Heme-Containing Peptides Using Non-Activated Immobilized Metal Affinity Cchromatography Resin Enrichment and Higher-Energy Collisional Dissociation

    SciTech Connect (OSTI)

    Zhang, Haizhen; Yang, Feng; Qian, Weijun; Brown, Roslyn N.; Wang, Yuexi; Merkley, Eric D.; Park, Jea H.; Monroe, Matthew E.; Purvine, Samuel O.; Moore, Ronald J.; Shi, Liang; Fredrickson, Jim K.; Pasa-Tolic, Ljiljana; Smith, Richard D.; Lipton, Mary S.

    2011-10-01

    c-type cytochromes play essential roles in many biological activities of both prokaryotic and eukaryotic cells, including electron transfer, enzyme catalysis and induction of apoptosis. We report a novel enrichment strategy for identifying c-type heme-containing peptides that uses non-activated IMAC resin. The strategy demonstrated at least seven-fold enrichment for heme-containing peptides digested from a cytochrome c protein standard, and quantitative linear performance was also assessed for heme-containing peptide enrichment. Heme-containing peptides extracted from the periplasmic fraction of Shewanella oneidensis MR-1 were further identified using higher-energy collisional dissociation tandem mass spectrometry. The results demonstrated the applicability of this enrichment strategy to identify c-type heme-containing peptides from a highly complex biological sample, and at the same time, confirmed the periplasmic localization of heme-containing proteins during suboxic respiration activities of S. oneidensis MR-1.

  16. Blend Down Monitoring System Fissile Mass Flow Monitor and its Implementation at the Siberian Chemical Enterprise, Seversk, Russia

    SciTech Connect (OSTI)

    Uckan, T

    2005-07-28

    In this paper the implementation plans and preparations for installation of the Fissile Mass Flow Monitor (FMFM) equipment at the Siberian Chemical Enterprise (SChE), Seversk, Russia, are presented. The FMFM, developed by Oak Ridge National Laboratory, is part of the Blend Down Monitoring System (BDMS) for the U.S. Department of Energy Highly Enriched Uranium (HEU) Transparency Implementation Program. The BDMS provides confidence to the United States that the Russian nuclear facilities supplying the lower assay ({approx}4%) product low enriched uranium (PLEU) to the United States from down-blended weapon-grade HEU are meeting the nonproliferation goals of the government-to-government HEU purchase agreement signed between the Russian Federation and the United States in 1993. The first BDMS has been operational at Ural Electrochemical Integrated Plant, Novouralsk, since February 1999. The second BDMS has been operational at Electro Chemical Plant, Zelenogorsk, since March 2003. These systems are successfully providing HEU transparency data to the United States. The third BDMS was successfully installed on the HEU down-blending tee in the SChE Enrichment Plant in October 2004. The FMFM makes use of a set of thermalized {sup 252}Cf spontaneous neutron sources for modulated fission activation of the UF{sub 6} gas stream for measuring the {sup 235}U fissile mass flow rate. To do this, the FMFM measures the transport time of the fission fragments created from the fission activation process under the modulated source to the downstream detectors by detecting the delayed gamma rays from the fission fragments retained in the flow. The FMFM provides unattended nonintrusive measurements of the {sup 235}U mass flow of the UF{sub 6} gas in the blending tee legs of HEU, the LEU blend stock, and the resulting P-LEU. The FMFM also confirms that highly enriched UF{sub 6} gas identified in the HEU leg flows through the blending tee into the P-LEU leg. This report contains details of the SChE FMFM equipment characteristics as well as the technical installation requirements and the latest measurement results.

  17. Method for production of an isotopically enriched compound

    DOE Patents [OSTI]

    Watrous, Matthew G.

    2012-12-11

    A method is presented for producing and isolating an isotopically enriched compound of a desired isotope from a parent radionuclide. The method includes forming, or placing, a precipitate containing a parent radionuclide of the desired daughter isotope in a first reaction zone and allowing sufficient time for the parent to decay into the desired gaseous daughter radioisotope. The method further contemplates collecting the desired daughter isotope as a solid in a second reaction zone through the application of temperatures below the freezing point of the desired isotope to a second reaction zone that is connected to the first reaction zone. Specifically, a method is presented for producing isotopically enriched compounds of xenon, including the radioactive isotope Xe-131m and the stable isotope Xe-131.

  18. Natural uranium/conversion services/enrichment services

    SciTech Connect (OSTI)

    1993-12-31

    This article is the 1993 uranium market summary. During this reporting period, there were 50 deals in the concentrates market, 26 deals in the UF6 market, and 14 deals for enrichment services. In the concentrates market, the restricted value closed $0.15 higher at $9.85, and the unrestricted value closed down $0.65 at $7.00. In the UF6 market, restricted prices fluctuated and closed higher at $31.00, and unrestricted prices closed at their initial value of $24.75. The restricted transaction value closed at $10.25 and the unrestricted value closed at $7.15. In the enrichment services market, the restricted value moved steadily higher to close at $84.00 per SWU, and the unrestricted value closed at its initial value of $68.00 per SWU.

  19. Comments on proposed legislation to restructure DOE's uranium enrichment program

    SciTech Connect (OSTI)

    Not Available

    1991-04-01

    This book focuses on H.R.145, H.R.788, and S.210. Each of the proposed bills would restructure DOE's enrichment program as a government corporation with private financing and would encourage the eventual sale of the corporation to the private sector. In doing so, the bills would, among other things, allow the corporation to set prices to maximize long-term returns; establish a fund to meet the costs of decontamination, decommissioning, and other environmental cleanup costs associated with uranium enrichment activities; transfer interest in DOE's new atomic vapor laser isotope separation (AVLIS) process to the new corporation; and, except for H.R. 145, require the government to pay its share of the costs to clean up mill tailings (mining wastes) generated under government contracts.

  20. Simulation of transportation of low enriched uranium solutions

    SciTech Connect (OSTI)

    Hope, E.P.; Ades, M.J.

    1996-08-01

    A simulation of the transportation by truck of low enriched uranium solutions has been completed for NEPA purposes at the Savannah River Site. The analysis involves three distinct source terms, and establishes the radiological risks of shipment to three possible destinations. Additionally, loading accidents were analyzed to determine the radiological consequences of mishaps during handling and delivery. Source terms were developed from laboratory measurements of chemical samples from low enriched uranium feed materials being stored at SRS facilities, and from manufacturer data on transport containers. The transportation simulations were accomplished over the INTERNET using the DOE TRANSNET system at Sandia National Laboratory. The HIGHWAY 3.3 code was used to analyze routing scenarios, and the RADTRAN 4 code was used to analyze incident free and accident risks of transporting radiological materials. Loading accidents were assessed using the Savannah River Site AXAIR89Q and RELEASE 2 codes.

  1. News Media invited to interview Jefferson Lab summer science enrichment

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    program participants; cover closing Poster Session | Jefferson Lab SULI 2004 participant Rachel Black, SULI 2004 participant, talks to Alan Gavalya (far left), Physics Division, and Jim Clark, Accelerator Division, about her work with JLab's Detector Group. Photo: Greg Adams, JLab Media Advisory: News Media invited to interview Jefferson Lab summer science enrichment program participants; cover closing Poster Session July 29, 2005 News Media representatives are invited to interview,

  2. Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR

    SciTech Connect (OSTI)

    Hanson, A.L.; Diamond, D.

    2011-09-30

    A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Previously, the design of the LEU fuel had been determined in order to provide the users of the NBSR with the same cycle length as exists for the current HEU fueled reactor. The fuel composition at different points within an equilibrium fuel cycle had also been determined. In the present study, neutronics parameters have been calculated for these times in the fuel cycle for both the existing HEU and the proposed LEU equilibrium cores. The results showed differences between the HEU and LEU cores that would not lead to any significant changes in the safety analysis for the converted core. In general the changes were reasonable except that the figure-of-merit for neutrons that can be used by experimentalists shows there will be a 10% reduction in performance. The calculations included kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions.

  3. Future of the Department of Energy's uranium enrichment enterprise

    SciTech Connect (OSTI)

    Sewell, P.G.

    1991-11-01

    The national energy strategy (NES) developed at President Bush's direction provides a focus for the US Department of Energy (DOE) future policy and funding initiatives including those of the uranium enrichment enterprise. The NES identifies an important and continuing role for nuclear energy as part of a balanced array of energy sources for meeting US energy needs, especially the growing demand for electricity. For many years, growth in US electricity demand has exhibited a strong correlation with growth in gross national product. NEW projections indicate that the US will need between 190 and 275 GW of additional system capacity by 2010. In order to unable nuclear power to help meet this need, the NEW establishes basic objectives for nuclear power. These objectives are to have a first order of a new nuclear power plant by 1995 and to have such a plant operational by 2000. The expansion of nuclear power anticipated in the NEW affirms a continuing need for a strong domestic uranium enrichment services supply capability. In terms of the future outlook for uranium enrichment, the atomic vapor laser isotope separation (AVLIS) technology continues to hold great promise for commercial application. If AVLIS efforts are successful, significant financial benefits from the commercial use of AVLIS will be realized by customers and the AVLIS deployment entity by approximately the year 2000 and thereafter.

  4. EIS-0471: Areva Eagle Rock Enrichment Facility in Bonneville County, ID |

    Energy Savers [EERE]

    Department of Energy 1: Areva Eagle Rock Enrichment Facility in Bonneville County, ID EIS-0471: Areva Eagle Rock Enrichment Facility in Bonneville County, ID May 20, 2011 delete me old download page duplicate

  5. Jefferson Lab Offers Science Enrichment Program for 5th, 6th...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    3 Jefferson Lab Offers Science Enrichment Program for 5th, 6th & 8th Grade Teachers; ... its science enrichment program for fifth-, sixth- and eighth-grade teachers of science. ...

  6. Jefferson Lab Offers Science Enrichment Program for 5th, 6th...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Jefferson Lab Offers Science Enrichment Program for 5th, 6th & 8th Grade Teachers Teachers ... its science enrichment program for fifth-, sixth- and eighth-grade teachers of science. ...

  7. Jefferson Lab Offers Science Enrichment Program for 5th, 6th...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Jefferson Lab Offers Science Enrichment Program for 5th, 6th & 8th Grade Teachers; ... its science enrichment program for fifth-, sixth- and eighth-grade teachers of science. ...

  8. Centrifuge enrichment plants. (Latest citations from the NTIS data base). Published Search

    SciTech Connect (OSTI)

    Not Available

    1992-09-01

    The bibliography contains citations concerning the design, control, monitoring, and safety of centrifuge enrichment plants. Power supplies, enrichment plant safeguards, facility design, cascade heater test loops to monitor the enrichment process, inspection strategies, and the socioeconomic effects of centrifuge enrichment plants are examined. Radioactive waste disposal problems are briefly considered. (Contains a minimum of 169 citations and includes a subject term index and title list.)

  9. An Inspector's Assessment of the New Model Safeguards Approach for Enrichment Plants

    SciTech Connect (OSTI)

    Curtis, Michael M.

    2007-07-31

    This conference paper assesses the changes that are being made to the Model Safeguards Approach for Gas Centrifuge Enrichment Plants.

  10. Safeguards Verification Measurements using Laser Ablation, Absorbance Ratio Spectrometry in Gaseous Centrifuge Enrichment Plants

    SciTech Connect (OSTI)

    Anheier, Norman C.; Cannon, Bret D.; Qiao, Hong; Phillips, Jon R.

    2012-07-01

    Laser Ablation Absorbance Ratio Spectrometry (LAARS) is a new verification measurement technology under development at the US Department of Energy’s (DOE) Pacific Northwest National Laboratory (PNNL). LAARS uses three lasers to ablate and then measure the relative isotopic abundance of uranium compounds. An ablation laser is tightly focused on uranium-bearing solids producing a small plume containing uranium atoms. Two collinear wavelength-tuned spectrometry lasers transit through the plume and the absorbance of U-235 and U-238 isotopes are measured to determine U-235 enrichment. The measurement has high relative precision and detection limits approaching the femtogram range for uranium. It is independent of chemical form and degree of dilution with nuisance dust and other materials. High speed sample scanning and pinpoint characterization allow measurements on millions of particles/hour to detect and analyze the enrichment of trace uranium in samples. The spectrometer is assembled using commercially available components at comparatively low cost, and features a compact and low power design. Future designs can be engineered for reliable, autonomous deployment within an industrial plant environment. Two specific applications of the spectrometer are under development: 1) automated unattended aerosol sampling and analysis and 2) on-site small sample destructive assay measurement. The two applications propose game-changing technological advances in gaseous centrifuge enrichment plant (GCEP) safeguards verification. The aerosol measurement instrument, LAARS-environmental sampling (ES), collects aerosol particles from the plant environment in a purpose-built rotating drum impactor and then uses LAARS-ES to quickly scan the surface of the impactor to measure the enrichments of the captured particles. The current approach to plant misuse detection involves swipe sampling and offsite analysis. Though this approach is very robust it generally requires several months to obtain results from a given sample collection. The destructive assay instrument, LAARS-destructive assay (DA), uses a simple purpose-built fixture with a sampling planchet to collect adsorbed UF6 gas from a cylinder valve or from a process line tap or pigtail. A portable LAARS-DA instrument scans the microgram quantity of uranium collected on the planchet and the assay of the uranium is measured to ~0.15% relative precision. Currently, destructive assay samples for bias defect measurements are collected in small sample cylinders for offsite mass spectrometry measurement.

  11. FEMO, A FLOW AND ENRICHMENT MONITOR FOR VERIFYING COMPLIANCE WITH INTERNATIONAL SAFEGUARDS REQUIREMENTS AT A GAS CENTRIFUGE ENRICHMENT FACILITY

    SciTech Connect (OSTI)

    Gunning, John E; Laughter, Mark D; March-Leuba, Jose A

    2008-01-01

    A number of countries have received construction licenses or are contemplating the construction of large-capacity gas centrifuge enrichment plants (GCEPs). The capability to independently verify nuclear material flows is a key component of international safeguards approaches, and the IAEA does not currently have an approved method to continuously monitor the mass flow of 235U in uranium hexafluoride (UF6) gas streams. Oak Ridge National Laboratory is investigating the development of a flow and enrichment monitor, or FEMO, based on an existing blend-down monitoring system (BDMS). The BDMS was designed to continuously monitor both 235U mass flow and enrichment of UF6 streams at the low pressures similar to those which exists at GCEPs. BDMSs have been installed at three sites-the first unit has operated successfully in an unattended environment for approximately 10 years. To be acceptable to GCEP operators, it is essential that the instrument be installed and maintained without interrupting operations. A means to continuously verify flow as is proposed by FEMO will likely be needed to monitor safeguards at large-capacity plants. This will enable the safeguards effectiveness that currently exists at smaller plants to be maintained at the larger facilities and also has the potential to reduce labor costs associated with inspections at current and future plants. This paper describes the FEMO design requirements, operating capabilities, and development work required before field demonstration.

  12. Application of cadmium-zinc-telluride detectors in U-235 enrichment measurements

    SciTech Connect (OSTI)

    Ruhter, W.D.; Gunnink, R.

    1994-04-01

    High-resolution, gamma- and x-ray spectrometry are used routinely in nuclear safeguards verification measurements of plutonium and uranium in the field. These measurements are mostly performed with high-purity germanium (HPGe) detectors, that require cooling to liquid-nitrogen temperatures, thus limiting their utility in field and unattended safeguards measurement applications. NaI scintillation detectors do not require cooling, but their moderate energy resolution (10% at 122 keV) is insufficient in most cases for reliable verification measurements. Semiconductor detectors that operate at room temperature, such as cadmium-zinc-telluride (CdZnTe) detectors, with energy resolution performance reaching 2.0% at 122 keV may complement HPGe detectors for certain safeguards verification applications. The authors used a 5x5x5 mm CdZnTe detector to measure U-235 enrichments ranging from 3% to 75%. They use a spectrum analysis technique that fits U-235, U-238, and U K x-ray response profiles to data in the 89- to 100-keV region of gamma-ray spectrum. From the relative magnitudes of the U-235 and U-238 profiles they determine the U-235 enrichment with an accuracy of about 10% with CdZnTe detectors.

  13. SR0601

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    May 4, 2006 Media Contact: James Giusti (803) 952-7697 SRS Makes 200th Shipment of Low Enriched Uranium AIKEN, S.C. -- A joint National Nuclear Security Administration and Office of Environmental Management nonproliferation program at the Savannah River Site (SRS) completed its 200th shipment of low enriched uranium (LEU) to Erwin, TN, earlier today. This is the final step at SRS in the process of turning weapons-grade uranium into electricity. Under SRS' Highly Enriched Uranium (HEU) Blend Down

  14. Compton DIV: Using a Compton-Based Gamma-Ray Imager for Design Information Verification of Uranium Enrichment Plants

    SciTech Connect (OSTI)

    Burks, M; Verbeke, J; Dougan, A; Wang, T; Decman, D

    2009-07-04

    A feasibility study has been performed to determine the potential usefulness of Compton imaging as a tool for design information verification (DIV) of uranium enrichment plants. Compton imaging is a method of gamma-ray imaging capable of imaging with a 360-degree field of view over a broad range of energies. These systems can image a room (with a time span on the order of one hour) and return a picture of the distribution and composition of radioactive material in that room. The effectiveness of Compton imaging depends on the sensitivity and resolution of the instrument as well the strength and energy of the radioactive material to be imaged. This study combined measurements and simulations to examine the specific issue of UF{sub 6} gas flow in pipes, at various enrichment levels, as well as hold-up resulting from the accumulation of enriched material in those pipes. It was found that current generation imagers could image pipes carrying UF{sub 6} in less than one hour at moderate to high enrichment. Pipes with low enriched gas would require more time. It was also found that hold-up was more amenable to this technique and could be imaged in gram quantities in a fraction of an hour. another questions arises regarding the ability to separately image two pipes spaced closely together. This depends on the capabilities of the instrument in question. These results are described in detail. In addition, suggestions are given as to how to develop Compton imaging as a tool for DIV.

  15. Uranium mineralization in fluorine-enriched volcanic rocks

    SciTech Connect (OSTI)

    Burt, D.M.; Sheridan, M.F.; Bikun, J.; Christiansen, E.; Correa, B.; Murphy, B.; Self, S.

    1980-09-01

    Several uranium and other lithophile element deposits are located within or adjacent to small middle to late Cenozoic, fluorine-rich rhyolitic dome complexes. Examples studied include Spor Mountain, Utah (Be-U-F), the Honeycomb Hills, Utah (Be-U), the Wah Wah Mountains, Utah (U-F), and the Black Range-Sierra Cuchillo, New Mexico (Sn-Be-W-F). The formation of these and similar deposits begins with the emplacement of a rhyolitic magma, enriched in lithophile metals and complexing fluorine, that rises to a shallow crustal level, where its roof zone may become further enriched in volatiles and the ore elements. During initial explosive volcanic activity, aprons of lithicrich tuffs are erupted around the vents. These early pyroclastic deposits commonly host the mineralization, due to their initial enrichment in the lithophile elements, their permeability, and the reactivity of their foreign lithic inclusions (particularly carbonate rocks). The pyroclastics are capped and preserved by thick topaz rhyolite domes and flows that can serve as a source of heat and of additional quantities of ore elements. Devitrification, vapor-phase crystallization, or fumarolic alteration may free the ore elements from the glassy matrix and place them in a form readily leached by percolating meteoric waters. Heat from the rhyolitic sheets drives such waters through the system, generally into and up the vents and out through the early tuffs. Secondary alteration zones (K-feldspar, sericite, silica, clays, fluorite, carbonate, and zeolites) and economic mineral concentrations may form in response to this low temperature (less than 200 C) circulation. After cooling, meteoric water continues to migrate through the system, modifying the distribution and concentration of the ore elements (especially uranium).

  16. Using Process Load Cell Information for IAEA Safeguards at Enrichment Plants

    SciTech Connect (OSTI)

    Laughter, Mark D; Whitaker, J Michael; Howell, John

    2010-01-01

    Uranium enrichment service providers are expanding existing enrichment plants and constructing new facilities to meet demands resulting from the shutdown of gaseous diffusion plants, the completion of the U.S.-Russia highly enriched uranium downblending program, and the projected global renaissance in nuclear power. The International Atomic Energy Agency (IAEA) conducts verification inspections at safeguarded facilities to provide assurance that signatory States comply with their treaty obligations to use nuclear materials only for peaceful purposes. Continuous, unattended monitoring of load cells in UF{sub 6} feed/withdrawal stations can provide safeguards-relevant process information to make existing safeguards approaches more efficient and effective and enable novel safeguards concepts such as information-driven inspections. The IAEA has indicated that process load cell monitoring will play a central role in future safeguards approaches for large-scale gas centrifuge enrichment plants. This presentation will discuss previous work and future plans related to continuous load cell monitoring, including: (1) algorithms for automated analysis of load cell data, including filtering methods to determine significant weights and eliminate irrelevant impulses; (2) development of metrics for declaration verification and off-normal operation detection ('cylinder counting,' near-real-time mass balancing, F/P/T ratios, etc.); (3) requirements to specify what potentially sensitive data is safeguards relevant, at what point the IAEA gains on-site custody of the data, and what portion of that data can be transmitted off-site; (4) authentication, secure on-site storage, and secure transmission of load cell data; (5) data processing and remote monitoring schemes to control access to sensitive and proprietary information; (6) integration of process load cell data in a layered safeguards approach with cross-check verification; (7) process mock-ups constructed to provide simulated load cell data; (8) hardware and software implementation for process load cell data collection; (9) costs associated with unattended monitoring of load cells (for both operator and inspector) weighed against the potential benefits of having access to such data; (10) results from field tests of load cell data collection systems in operating facilities; and (11) use of unattended load cell data to increase efficiency of on-site inspection schedules and activities.

  17. Separation of carbon nanotubes into chirally enriched fractions

    DOE Patents [OSTI]

    Doorn, Stephen K.; Niyogi, Sandip

    2012-04-10

    A mixture of single-walled carbon nanotubes ("SWNTs") is separated into fractions of enriched chirality by preparing an aqueous suspension of a mixture of SWNTs and a surfactant, injecting a portion of the suspension on a column of separation medium having a density gradient, and centrifuging the column. In some embodiments, salt is added prior to centrifugation. In other embodiments, the centrifugation is performed at a temperature below room temperature. Fractions separate as colored bands in the column. The diameter of the separated SWNTs decreases with increasing density along the gradient of the column. The colored bands can be withdrawn separately from the column.

  18. New Prototype Safeguards Technology Offers Improved Confidence and Automation for Uranium Enrichment Facilities

    SciTech Connect (OSTI)

    Brim, Cornelia P.

    2013-04-01

    An important requirement for the international safeguards community is the ability to determine the enrichment level of uranium in gas centrifuge enrichment plants and nuclear fuel fabrication facilities. This is essential to ensure that countries with nuclear nonproliferation commitments, such as States Party to the Nuclear Nonproliferation Treaty, are adhering to their obligations. However, current technologies to verify the uranium enrichment level in gas centrifuge enrichment plants or nuclear fuel fabrication facilities are technically challenging and resource-intensive. NNSA’s Office of Nonproliferation and International Security (NIS) supports the development, testing, and evaluation of future systems that will strengthen and sustain U.S. safeguards and security capabilities—in this case, by automating the monitoring of uranium enrichment in the entire inventory of a fuel fabrication facility. One such system is HEVA—hybrid enrichment verification array. This prototype was developed to provide an automated, nondestructive assay verification technology for uranium hexafluoride (UF6) cylinders at enrichment plants.

  19. copy

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    convert program Material Management and Minimization The Office of Material Management and Minimization (M3) presents an integrated approach to addressing the persistent threat posed by nuclear materials through a full cycle of materials management and minimization efforts. Consistent with the President's highly enriched uranium (HEU) and... Conversion A key starting point for material management and minimization is reducing the civilian use of and demands for weapon-grade nuclear material. The

  20. Audit Report: OAS-L-12-07 | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    7 Audit Report: OAS-L-12-07 July 20, 2012 The Global Threat Reduction Initiative's Molybdenum-99 Program Molybdenum-99 (Mo-99) is used in the production of technetium-99m, the most commonly used medical radioisotope in the world. Because the U.S. lacks a domestic production capability, its demand is met by other countries, whose processes have recently proven unreliable. In addition, the foreign producers utilize highly enriched uranium (HEU), a practice contrary to the National Nuclear Security

  1. The Secretary of Energy

    National Nuclear Security Administration (NNSA)

    Secretary of Energy Washington, DC 20585 DETERMINATION OF COMPLETION OF THE RUSSIAN HEU AGREEMENT FOR PURPOSES OF THE USEC PRIVATIZATION ACT OF 1996, AS AMENDED Having considered the provisions of the USEC Privatization Act of 1996, as amended, (42 U.S.C.§ 2297h, et seq.), and the requirements for completion of the Agreement Between the Government of the United States of America and the Government of the Russian Federation Concerning the Disposition of Highly- Enriched Uranium Extracted From

  2. material protection | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    material protection Material Management and Minimization The Office of Material Management and Minimization (M3) presents an integrated approach to addressing the persistent threat posed by nuclear materials through a full cycle of materials management and minimization efforts. Consistent with the President's highly enriched uranium (HEU) and... Nonproliferation Working in close collaboration with DOE laboratories, DNN develops and tests new technologies to advance U.S. capabilities to monitor

  3. material recovery | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    recovery Material Management and Minimization The Office of Material Management and Minimization (M3) presents an integrated approach to addressing the persistent threat posed by nuclear materials through a full cycle of materials management and minimization efforts. Consistent with the President's highly enriched uranium (HEU) and... Nonproliferation Working in close collaboration with DOE laboratories, DNN develops and tests new technologies to advance U.S. capabilities to monitor

  4. Joint Statement by President Obama and Prime Minister Elio Di Rupo of

    National Nuclear Security Administration (NNSA)

    Belgium on the 2014 Nuclear Security Summit | National Nuclear Security Administration President Obama and Prime Minister Elio Di Rupo of Belgium on the 2014 Nuclear Security Summit March 24, 2014 See a fact sheet here. The White House Office of the Press Secretary Belgium and the United States of America are pleased to announce that they have jointly completed the removal of a significant amount of excess highly enriched uranium (HEU) and separated plutonium from Belgium. At the 2012

  5. Dynamic System Simulation of the KRUSTY Experiment (Technical Report) |

    Office of Scientific and Technical Information (OSTI)

    SciTech Connect SciTech Connect Search Results Technical Report: Dynamic System Simulation of the KRUSTY Experiment Citation Details In-Document Search Title: Dynamic System Simulation of the KRUSTY Experiment The proposed KRUSTY experiment is a demonstration of a reactor operating at power. The planned experimental configuration includes a highly enriched uranium (HEU) reflected core, cooled by multiple heat pipes leading to Stirling engines for primary heat rejection. Operating power is

  6. Testimony before the Senate Committee on Energy & Natural Resources |

    National Nuclear Security Administration (NNSA)

    National Nuclear Security Administration Testimony before the Senate Committee on Energy & Natural Resources February 01, 2011 Chairman Bingaman, Ranking Member Murkowski, and Committee Members, thank you for the opportunity to testify about the National Nuclear Security Administration's (NNSA's) support for accelerating development of a domestic commercial supply of Molybdenum-99 (Mo-99) without using highly enriched uranium (HEU). This effort is part of our larger global

  7. EIS-0218-SA-07: Supplement Analysis | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    8-SA-07: Supplement Analysis EIS-0218-SA-07: Supplement Analysis Foreign Research Reactor Spent Nuclear Fuel Acceptance Program: Highly Enriched Uranium Target Residue Material Transportation This Supplement Analysis (SA) summarizes recent activities related to the safe transport of target residue material (TRM), which contains HEU of U. S. origin, from Canada as part of DOE's foreign research reactor (FRR) acceptance program. The material would be transported in liquid form to SRS. It would be

  8. Belgium | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Belgium Joint Statement by President Obama and Prime Minister Elio Di Rupo of Belgium on the 2014 Nuclear Security Summit See a fact sheet here. The White HouseOffice of the Press Secretary Belgium and the United States of America are pleased to announce that they have jointly completed the removal of a significant amount of excess highly enriched uranium (HEU) and separated plutonium from Belgium. At

  9. NNSA Awards Mo-99 Cooperative Agreement to General Atomics | National

    National Nuclear Security Administration (NNSA)

    Nuclear Security Administration Library / Press Releases NNSA Awards Mo-99 Cooperative Agreement to General Atomics September 30, 2015 WASHINGTON, DC - Today, the Department of Energy's National Nuclear Security Administration (DOE/NNSA) announced that it will award a cooperative agreement to General Atomics (GA) to support its project for domestic production of molybdenum-99 (Mo-99) without highly enriched uranium (HEU). Mo-99 is the parent isotope of technetium-99m, which is the most

  10. Nucleosynthesis in helium-enriched asymptotic giant branch models: Implications for heavy element enrichment in ω Centauri

    SciTech Connect (OSTI)

    Karakas, Amanda I.; Marino, Anna F.; Nataf, David M.

    2014-03-20

    We investigate the effect of helium enrichment on the evolution and nucleosynthesis of low-mass asymptotic giant branch (AGB) stars of 1.7 M {sub ☉} and 2.36 M {sub ☉} with a metallicity of Z = 0.0006 ([Fe/H] ≈–1.4). We calculate evolutionary sequences with the primordial helium abundance (Y = 0.24) and with helium-enriched compositions (Y = 0.30, 0.35, 0.40). For comparison, we calculate models of the same mass but at a lower metallicity Z = 0.0003 ([Fe/H] ≈–1.8) with Y = 0.24. Post-processing nucleosynthesis calculations are performed on each of the evolutionary sequences to determine the production of elements from hydrogen to bismuth. Elemental surface abundance predictions and stellar yields are presented for each model. The models with enriched helium have shorter main sequence and AGB lifetimes, and they enter the AGB with a more massive hydrogen-exhausted core than the primordial helium model. The main consequences are as follows: (1) low-mass AGB models with enhanced helium will evolve more than twice as fast, giving them the chance to contribute sooner to the chemical evolution of the forming globular clusters, and (2) the stellar yields will be strongly reduced relative to their primordial helium counterparts. An increase of ΔY = 0.10 at a given mass decreases the yields of carbon by up to ≈60% and of fluorine by up to 80%; it also decreases the yields of the s-process elements barium and lanthanum by ≈45%. While the yields of first s-process peak elements strontium, yttrium, and zirconium decrease by up to 50%, the yields of rubidium either do not change or increase.

  11. MAVTgsa: An R Package for Gene Set (Enrichment) Analysis

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Chien, Chih-Yi; Chang, Ching-Wei; Tsai, Chen-An; Chen, James J.

    2014-01-01

    Gene semore » t analysis methods aim to determine whether an a priori defined set of genes shows statistically significant difference in expression on either categorical or continuous outcomes. Although many methods for gene set analysis have been proposed, a systematic analysis tool for identification of different types of gene set significance modules has not been developed previously. This work presents an R package, called MAVTgsa, which includes three different methods for integrated gene set enrichment analysis. (1) The one-sided OLS (ordinary least squares) test detects coordinated changes of genes in gene set in one direction, either up- or downregulation. (2) The two-sided MANOVA (multivariate analysis variance) detects changes both up- and downregulation for studying two or more experimental conditions. (3) A random forests-based procedure is to identify gene sets that can accurately predict samples from different experimental conditions or are associated with the continuous phenotypes. MAVTgsa computes the P values and FDR (false discovery rate) q -value for all gene sets in the study. Furthermore, MAVTgsa provides several visualization outputs to support and interpret the enrichment results. This package is available online.« less

  12. Isotopically Enriched Films and Nanostructures by Ultrafast Pulsed Laser Deposition

    SciTech Connect (OSTI)

    Peter Pronko

    2004-12-13

    This project involved a systematic study to apply newly discovered isotopic enrichment effects in laser ablation plumes to the fabrication of isotopically engineered thin films, superlattices, and nanostructures. The approach to this program involved using ultrafast lasers as a method for generating ablated plasmas that have preferentially structured isotopic content in the body of the ablation plasma plumes. In examining these results we have attempted to interpret the observations in terms of a plasma centrifuge process that is driven by the internal electro-magnetic fields of the plasma itself. The research plan involved studying the following phenomena in regard to the ablation plume and the isotopic mass distribution within it: (1) Test basic equations of steady state centrifugal motion in the ablation plasma. (2) Investigate angular distribution of ions in the ablation plasmas. (3) Examine interactions of plasma ions with self-generated magnetic fields. (3) Investigate ion to neutral ratios in the ablation plasmas. (5) Test concepts of plasma pumping. (6) Fabricate isotopically enriched nanostructures.

  13. LISSAT Analysis of a Generic Centrifuge Enrichment Plant

    SciTech Connect (OSTI)

    Lambert, H; Elayat, H A; O?Connell, W J; Szytel, L; Dreicer, M

    2007-05-31

    The U.S. Department of Energy (DOE) is interested in developing tools and methods for use in designing and evaluating safeguards systems for current and future plants in the nuclear power fuel cycle. The DOE is engaging several DOE National Laboratories in efforts applied to safeguards for chemical conversion plants and gaseous centrifuge enrichment plants. As part of the development, Lawrence Livermore National Laboratory has developed an integrated safeguards system analysis tool (LISSAT). This tool provides modeling and analysis of facility and safeguards operations, generation of diversion paths, and evaluation of safeguards system effectiveness. The constituent elements of diversion scenarios, including material extraction and concealment measures, are structured using directed graphs (digraphs) and fault trees. Statistical analysis evaluates the effectiveness of measurement verification plans and randomly timed inspections. Time domain simulations analyze significant scenarios, especially those involving alternate time ordering of events or issues of timeliness. Such simulations can provide additional information to the fault tree analysis and can help identify the range of normal operations and, by extension, identify additional plant operational signatures of diversions. LISSAT analyses can be used to compare the diversion-detection probabilities for individual safeguards technologies and to inform overall strategy implementations for present and future plants. Additionally, LISSAT can be the basis for a rigorous cost-effectiveness analysis of safeguards and design options. This paper will describe the results of a LISSAT analysis of a generic centrifuge enrichment plant. The paper will describe the diversion scenarios analyzed and the effectiveness of various safeguards systems alternatives.

  14. Process for preparing a chemical compound enriched in isotope content

    DOE Patents [OSTI]

    Michaels, Edward D.

    1982-01-01

    A process to prepare a chemical enriched in isotope content which includes: (a) A chemical exchange reaction between a first and second compound which yields an isotopically enriched first compound and an isotopically depleted second compound; (b) the removal of a portion of the first compound as product and the removal of a portion of the second compound as spent material; (c) the conversion of the remainder of the first compound to the second compound for reflux at the product end of the chemical exchange reaction region; (d) the conversion of the remainder of the second compound to the first compound for reflux at the spent material end of the chemical exchange region; and the cycling of the additional chemicals produced by one conversion reaction to the other conversion reaction, for consumption therein. One of the conversion reactions is an oxidation reaction, and the energy that it yields is used to drive the other conversion reaction, a reduction. The reduction reaction is carried out in a solid polymer electrolyte electrolytic reactor. The overall process is energy efficient and yields no waste by-products.

  15. Fissile Flow and Enrichment Monitor for GCEP Advanced Safeguards Application

    SciTech Connect (OSTI)

    March-Leuba, Jose A; Uckan, Taner

    2010-01-01

    This paper presents experimental data that demonstrate a concept for a {sup 235}U flow and enrichment monitor (FEMO) based on passive measurements of process equipment in gaseous centrifuge enrichment plants (GCEPs). The primary goal of the FEMO is to prevent, without using pipe penetrations or active interrogation with external sources, the production and diversion of undeclared nuclear material. This FEMO concept utilizes: (1) calibrated measurements of {sup 235}U density in cascade headers, and (2) measurements of pump inlet pressure and volumetric flow rate, which are correlated to the electrical power consumed by the GCEP pumps that transport UF{sub 6} from the cascade to the condensation cylinders. The {sup 235}U density is measured by counting 186 keV emissions using a NaI gamma detector located upstream of the pump. The pump inlet pressure and volumetric flow rate are determined using a correlation that is a function of the measured pump operational parameters (e.g., electric power consumption and rotational frequency) and the pumping configuration. The concept has been demonstrated in a low-pressure flow loop at Oak Ridge National Laboratory.

  16. Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR using a U7Mo Dispersion Fuel

    SciTech Connect (OSTI)

    Hanson A. L.; Diamond D.

    2014-06-30

    A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The LEU fuel may be a monolithic foil (LEUm) of U10Mo (10% molybdenum by weight in an alloy with uranium) or a dispersion of U7Mo in aluminum (LEUd). A previous report provided neutronic calculations for the LEUm fuel and this report presents the neutronics parameters for the LEUd fuel. The neutronics parameters for the LEUd fuel are compared to those previously obtained for the present HEU fuel and the proposed LEUm fuel. The results show no significant differences between the LEUm and the LEUd other than the LEUd fuel requires slightly less uranium than the LEUm fuel due to less molybdenum being present. The calculations include kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions under normal operation and with misloaded fuel elements.

  17. Certification of U.S. instrumentation in Russian nuclear processing facilities

    SciTech Connect (OSTI)

    D.H. Powell; J.N. Sumner

    2000-07-12

    Agreements between the United States (U.S.) and the Russian Federation (R.F.) require the down-blending of highly enriched uranium (HEU) from dismantled Russian Federation nuclear weapons. The Blend Down Monitoring System (BDMS) was jointly developed by the Los Alamos National Laboratory (LANL) and the Oak Ridge National Laboratory (ORNL) to continuously monitor the enrichments and flow rates in the HEU blending operations at the R.F. facilities. A significant requirement of the implementation of the BDMS equipment in R.F. facilities concerned the certification of the BDMS equipment for use in a Russian nuclear facility. This paper discusses the certification of the BDMS for installation in R.F. facilities, and summarizes the lessons learned from the process that can be applied to the installation of other U.S. equipment in Russian nuclear facilities.

  18. Impact of HFIR LEU Conversion on Beryllium Reflector Degradation Factors

    SciTech Connect (OSTI)

    Ilas, Dan

    2013-10-01

    An assessment of the impact of low enriched uranium (LEU) conversion on the factors that may cause the degradation of the beryllium reflector is performed for the High Flux Isotope Reactor (HFIR). The computational methods, models, and tools, comparisons with previous work, along with the results obtained are documented and discussed in this report. The report documents the results for the gas and neutronic poison production, and the heating in the beryllium reflector for both the highly enriched uranium (HEU) and LEU HFIR configurations, and discusses the impact that the conversion to LEU may have on these quantities. A time-averaging procedure was developed to calculate the isotopic (gas and poisons) production in reflector. The sensitivity of this approach to different approximations is gauged and documented. The results show that the gas is produced in the beryllium reflector at a total rate of 0.304 g/cycle for the HEU configuration; this rate increases by ~12% for the LEU case. The total tritium production rate in reflector is 0.098 g/cycle for the HEU core and approximately 11% higher for the LEU core. A significant increase (up to ~25%) in the neutronic poisons production in the reflector during the operation cycles is observed for the LEU core, compared to the HEU case, for regions close to the core s horizontal midplane. The poisoning level of the reflector may increase by more than two orders of magnitude during long periods of downtime. The heating rate in the reflector is estimated to be approximately 20% lower for the LEU core than for the HEU core. The decrease is due to a significantly lower contribution of the heating produced by the gamma radiation for the LEU core. Both the isotopic (gas and neutronic poisons) production and the heating rates are spatially non-uniform throughout the beryllium reflector volume. The maximum values typically occur in the removable reflector and close to the midplane.

  19. Variable oxygen/nitrogen enriched intake air system for internal combustion engine applications

    DOE Patents [OSTI]

    Poola, Ramesh B.; Sekar, Ramanujam R.; Cole, Roger L.

    1997-01-01

    An air supply control system for selectively supplying ambient air, oxygen enriched air and nitrogen enriched air to an intake of an internal combustion engine includes an air mixing chamber that is in fluid communication with the air intake. At least a portion of the ambient air flowing to the mixing chamber is selectively diverted through a secondary path that includes a selectively permeable air separating membrane device due a differential pressure established across the air separating membrane. The permeable membrane device separates a portion of the nitrogen in the ambient air so that oxygen enriched air (permeate) and nitrogen enriched air (retentate) are produced. The oxygen enriched air and the nitrogen enriched air can be selectively supplied to the mixing chamber or expelled to atmosphere. Alternatively, a portion of the nitrogen enriched air can be supplied through another control valve to a monatomic-nitrogen plasma generator device so that atomic nitrogen produced from the nitrogen enriched air can be then injected into the exhaust of the engine. The oxygen enriched air or the nitrogen enriched air becomes mixed with the ambient air in the mixing chamber and then the mixed air is supplied to the intake of the engine. As a result, the air being supplied to the intake of the engine can be regulated with respect to the concentration of oxygen and/or nitrogen.

  20. Monte Carlo modeling and analyses of YALINA booster subcritical assembly, Part III : low enriched uranium conversion analyses.

    SciTech Connect (OSTI)

    Talamo, A.; Gohar, Y. (Nuclear Engineering Division) [Nuclear Engineering Division

    2011-05-12

    This study investigates the performance of the YALINA Booster subcritical assembly, located in Belarus, during operation with high (90%), medium (36%), and low (21%) enriched uranium fuels in the assembly's fast zone. The YALINA Booster is a zero-power, subcritical assembly driven by a conventional neutron generator. It was constructed for the purpose of investigating the static and dynamic neutronics properties of accelerator driven subcritical systems, and to serve as a fast neutron source for investigating the properties of nuclear reactions, in particular transmutation reactions involving minor-actinides. The first part of this study analyzes the assembly's performance with several fuel types. The MCNPX and MONK Monte Carlo codes were used to determine effective and source neutron multiplication factors, effective delayed neutron fraction, prompt neutron lifetime, neutron flux profiles and spectra, and neutron reaction rates produced from the use of three neutron sources: californium, deuterium-deuterium, and deuterium-tritium. In the latter two cases, the external neutron source operates in pulsed mode. The results discussed in the first part of this report show that the use of low enriched fuel in the fast zone of the assembly diminishes neutron multiplication. Therefore, the discussion in the second part of the report focuses on finding alternative fuel loading configurations that enhance neutron multiplication while using low enriched uranium fuel. It was found that arranging the interface absorber between the fast and the thermal zones in a circular rather than a square array is an effective method of operating the YALINA Booster subcritical assembly without downgrading neutron multiplication relative to the original value obtained with the use of the high enriched uranium fuels in the fast zone.

  1. Systems approach used in the Gas Centrifuge Enrichment Plant

    SciTech Connect (OSTI)

    Rooks, W.A. Jr.

    1982-01-01

    A requirement exists for effective and efficient transfer of technical knowledge from the design engineering team to the production work force. Performance-Based Training (PBT) is a systematic approach to the design, development, and implementation of technical training. This approach has been successfully used by the US Armed Forces, industry, and other organizations. The advantages of the PBT approach are: cost-effectiveness (lowest life-cycle training cost), learning effectiveness, reduced implementation time, and ease of administration. The PBT process comprises five distinctive and rigorous phases: Analysis of Job Performance, Design of Instructional Strategy, Development of Training Materials and Instructional Media, Validation of Materials and Media, and Implementation of the Instructional Program. Examples from the Gas Centrifuge Enrichment Plant (GCEP) are used to illustrate the application of PBT.

  2. Calculation of Kinetics Parameters for the NBSR

    SciTech Connect (OSTI)

    Hanson A. L.; Diamond D.

    2012-03-06

    The delayed neutron fraction and prompt neutron lifetime have been calculated at different times in the fuel cycle for the NBSR when fueled with both high-enriched uranium (HEU) and low-enriched uranium (LEU) fuel. The best-estimate values for both the delayed neutron fraction and the prompt neutron lifetime are the result of calculations using MCNP5-1.60 with the most recent ENDFB-VII evaluations. The best-estimate values for the total delayed neutron fraction from fission products are 0.00665 and 0.00661 for the HEU fueled core at startup and end-of-cycle, respectively. For the LEU fuel the best estimate values are 0.00650 and 0.00648 at startup and end-of-cycle, respectively. The present recommendations for the delayed neutron fractions from fission products are smaller than the value reported previously of 0.00726 for the HEU fuel. The best-estimate values for the contribution from photoneutrons will remain as 0.000316, independent of the fuel or time in the cycle.The values of the prompt neutron lifetime as calculated with MCNP5-1.60 are compared to values calculated with two other independent methods and the results are in reasonable agreement with each other. The recommended, conservative values of the neutron lifetime for the HEU fuel are 650 {micro}s and 750 {micro}s for the startup and end-of-cycle conditions, respectively. For LEU fuel the recommended, conservative values are 600 {micro}s and 700 {micro}s for the startup and end-of-cycle conditions, respectively. In all three calculations, the prompt neutron lifetime was determined to be longer for the end-of-cycle equilibrium condition when compared to the startup condition. The results of the three analyses were in agreement that the LEU fuel will exhibit a shorter prompt neutron lifetime when compared to the HEU fuel.

  3. Centrifuge enrichment plants. January 1970-October 1988 (Citations from the NTIS data base). Report for January 1970-October 1988

    SciTech Connect (OSTI)

    Not Available

    1988-11-01

    This bibliography contains citations concerning the design, control, monitoring, and safety of centrifuge enrichment plants. Power supplies, enrichment plant safeguards, facility design, cascade heater test loops to monitor the enrichment process, inspection strategies, and the socio-economic effects of centrifuge enrichment plants are examined. Radioactive waste disposal problems are briefly considered. (Contains 151 citations fully indexed and including a title list.)

  4. 10 CFR 830 Major Modification Determination for Advanced Test Reactor LEU Fuel Conversion

    SciTech Connect (OSTI)

    Boyd D. Christensen; Michael A. Lehto; Noel R. Duckwitz

    2012-05-01

    The Advanced Test Reactor (ATR), located in the ATR Complex of the Idaho National Laboratory (INL), was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. The ATR is fueled with high-enriched uranium (HEU) matrix (UAlx) in an aluminum sandwich plate cladding. The National Nuclear Security Administration Global Threat Reduction Initiative (GTRI) strategic mission includes efforts to reduce and protect vulnerable nuclear and radiological material at civilian sites around the world. Converting research reactors from using HEU to low-enriched uranium (LEU) was originally started in 1978 as the Reduced Enrichment for Research and Test Reactors (RERTR) Program under the U.S. Department of Energy (DOE) Office of Science. Within this strategic mission, GTRI has three goals that provide a comprehensive approach to achieving this mission: The first goal, the driver for the modification that is the subject of this determination, is to convert research reactors from using HEU to LEU. Thus the mission of the ATR LEU Fuel Conversion Project is to convert the ATR and Advanced Test Reactor Critical facility (ATRC) (two of the six U.S. High-Performance Research Reactors [HPRR]) to LEU fuel by 2017. The major modification criteria evaluation of the project pre-conceptual design identified several issues that lead to the conclusion that the project is a major modification.

  5. Simulated Real-World Energy Impacts of a Thermally Sensitive Powertrain Considering Viscous Losses and Enrichment (Presentation)

    SciTech Connect (OSTI)

    Wood, E.; Gonder, J.; Lopp, S.; Jehlik, F.

    2014-09-01

    It is widely understood that cold-temperature engine operation negatively impacts vehicle fuel use due to a combination of increased friction (high-viscosity engine oil) and temporary enrichment (accelerated catalyst heating). However, relatively little effort has been dedicated to thoroughly quantifying these impacts across a large number of driving cycles and ambient conditions. This work leverages high-quality dynamometer data collected at various ambient conditions to develop a modeling framework for quantifying engine cold-start fuel penalties over a wide array of real-world usage profiles. Additionally, mitigation strategies including energy retention and exhaust heat recovery are explored with benefits quantified for each approach.

  6. Method for enriching a middle isotope using vibration-vibration pumping

    DOE Patents [OSTI]

    Rich, Joseph W.; Homicz, Gregory F.; Bergman, Richard C.

    1989-01-01

    Method for producing isotopically enriched material by vibration-vibration excitation of gaseous molecules wherein a middle mass isotope of an isotopic mixture including lighter and heavier mass isotopes preferentially populates a higher vibrational mode and chemically reacts to provide a product in which it is enriched. The method can be used for vibration-vibration enrichment of .sup.17 O in a CO reactant mixture.

  7. Improving Power Production in Acetate-Fed Microbial Fuel Cells via Enrichment of Exoelectrogenic Organisms in Flow-Through Systems

    SciTech Connect (OSTI)

    Borole, Abhijeet P; Hamilton, Choo Yieng; Vishnivetskaya, Tatiana A; Leak, David; Andras, Calin

    2009-01-01

    An exoelectrogenic, biofilm-forming microbial consortium was enriched in an acetate-fed microbial fuel cell (MFC) using a flow-through anode coupled to an air-cathode. Multiple parameters known to improve MFC performance were integrated in one design including electrode spacing, specific electrode surface area, flow-through design, minimization of dead volume within anode chamber, and control of external resistance. In addition, continuous feeding of carbon source was employed and the MFC was operated at intermittent high flows to enable removal of non-biofilm forming organisms over a period of six months. The consortium enriched using the modified design and operating conditions resulted in a power density of 345 W m-3 of net anode volume (3650 mW m-2), when coupled to a ferricyanide cathode. The enriched consortium included -, -, -Proteobacteria, Bacteroidetes and Firmicutes. Members of the order Rhodocyclaceae and Burkholderiaceae (Azospira spp. (49%), Acidovorax spp. (11%) and Comamonas spp. (7%)), dominated the microbial consortium. Denaturing gradient gel electrophoresis (DGGE) analysis based on primers selective for Archaea suggested a very low abundance of methanogens. Limiting the delivery of the carbon source via continuous feeding corresponding to the maximum cathodic oxidation rates permitted in the flow-through, air-cathode MFC resulted in coulombic efficiencies reaching 88 5.7%.

  8. Deuterium enrichment by selective photoinduced dissociation of a multihalogenated organic compound

    DOE Patents [OSTI]

    Marling, John B.; Herman, Irving P.

    1981-01-01

    A method for deuterium enrichment by photoinduced dissociation which uses as the deuterium source a multihalogenated organic compound selected from the group consisting of a dihalomethane, a trihalomethane, a 1,2-dihaloethene, a trihaloethene, a tetrahaloethane and a pentahaloethane. The multihalogenated organic compound is subjected to intense infrared radiation at a preselected wavelength to selectively excite and thereby induce dissociation of substantially only those molecules containing deuterium to provide a deuterium enriched dissociation product. The deuterium enriched product may be combusted with oxygen to provide deuterium enriched water. The deuterium depleted undissociated molecules may be redeuterated by treatment with a deuterium source such as water.

  9. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  10. Status of the RERTR (Reduced Enrichment Research and Test Reactor) Program

    SciTech Connect (OSTI)

    Travelli, A.

    1988-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1987, the major events, findings and activities of 1988 are reviewed. The US Nuclear Regulatory Commission issued a formal and generic approval of the use of U3Si2-Al dispersion fuel in research and test reactors, with densities up to 4.8 g U/cmT. New significant findings from postirradiation examinations, from ion-beam irradiations, and from analytical modeling, have raised serious doubts about the potential of LEU U3Si-Al dispersion fuel for applications requiring very high uranium densities and high burnups (>6 g U/cmT, >50% burnup). As a result of these findings, the fuel development efforts have been redirected towards three new initiatives: (1) a systematic application of ion-beam irradiations to screen new materials; (2) application of Hot Isostatic Pressing (HIP) procedures to produce U3Si2-Al plates with high uranium densities and thin uniform cladding; and (3) application of HIP procedures to produce plates with U3Si wires imbedded in an aluminum matrix, achieving stability, high uranium density, and thin uniform cladding. The new fuel concepts hold the promise of extraordinary performance potential and require approximately five years to develop.

  11. Upgrading the HFIR Thermal-Hydraulic Legacy Code Using COMSOL

    SciTech Connect (OSTI)

    Bodey, Isaac T [ORNL] [ORNL; Arimilli, Rao V [ORNL] [ORNL; Freels, James D [ORNL] [ORNL

    2010-01-01

    Modernization of the High Flux Isotope Reactor (HFIR) thermal-hydraulic (TH) design and safety analysis capability is an important step in preparation for the conversion of the HFIR core from a high enriched uranium (HEU) fuel to a low enriched uranium (LEU) fuel. Currently, an important part of the HFIR TH analysis is based on the legacy Steady State Heat Transfer Code (SSHTC), which adds much conservatism to the safety analysis. The multi-dimensional multi-physics capabilities of the COMSOL environment allow the analyst to relax the number and magnitude of conservatisms, imposed by the SSHTC, to present a more physical model of the TH aspect of the HFIR.

  12. Upgrading the HFIR Thermal-Hydraulic Legacy Code Using COMSOL (Conference)

    Office of Scientific and Technical Information (OSTI)

    | SciTech Connect Upgrading the HFIR Thermal-Hydraulic Legacy Code Using COMSOL Citation Details In-Document Search Title: Upgrading the HFIR Thermal-Hydraulic Legacy Code Using COMSOL Modernization of the High Flux Isotope Reactor (HFIR) thermal-hydraulic (TH) design and safety analysis capability is an important step in preparation for the conversion of the HFIR core from a high enriched uranium (HEU) fuel to a low enriched uranium (LEU) fuel. Currently, an important part of the HFIR TH

  13. Characterization of the Mouse Brain Proteome Using Global Proteomic Analysis Complemented with Cysteinyl-Peptide Enrichment

    SciTech Connect (OSTI)

    Wang, Haixing H.; Qian, Weijun; Chin, Mark H.; Petyuk, Vladislav A.; Barry, Richard C.; Liu, Tao; Gritsenko, Marina A.; Mottaz, Heather M.; Moore, Ronald J.; Camp, David G.; Khan, Arshad H.; Smith, Desmond; Smith, Richard D.

    2006-02-01

    Given the growing interest in applying genomic and proteomic approaches for studying the mammalian brain using mouse models, we hereby present for the first time a comprehensive characterization of the mouse brain proteome. Preparation of the whole brain sample incorporated a highly efficient cysteinyl-peptide enrichment (CPE) technique to complement a global enzymatic digestion method. Both the global and the cysteinyl-enriched peptide samples were analyzed by SCX fractionation coupled with reversed phase LC-MS/MS analysis. A total of 48,328 different peptides were confidently identified (>98% confidence level), covering 7792 non-redundant proteins (~34% of the predicted mouse proteome). 1564 and 1859 proteins were identified exclusively from the cysteinyl-peptide and the global peptide samples, respectively, corresponding to 25% and 31% improvements in proteome coverage compared to analysis of only the global peptide or cysteinyl-peptide samples. The identified proteins provide a broad representation of the mouse proteome with little bias evident due to protein pI, molecular weight, and/or cellular localization. Approximately 26% of the identified proteins with gene ontology (GO) annotations were membrane proteins, with 1447 proteins predicted to have transmembrane domains, and many of the membrane proteins were found to be involved in transport and cell signaling. The MS/MS spectrum count information for the identified proteins was used to provide a measure of relative protein abundances. The mouse brain peptide/protein database generated from this study represents the most comprehensive proteome coverage for the mammalian brain to date, and the basis for future quantitative brain proteomic studies using mouse models.

  14. Environmental Survey preliminary report, Portsmouth Uranium Enrichment Complex, Piketon, Ohio

    SciTech Connect (OSTI)

    Not Available

    1987-08-01

    This report presents the preliminary findings from the first phase of the Environmental Survey of the United States Department of Energy (DOE) Portsmouth Uranium Enrichment Complex (PUEC), conducted August 4 through August 15, 1986. The Survey is being conducted by an interdisciplinary team of environmental specialists, led and managed by the Office of Environment, Safety and Health's Office of Environmental Audit. Team specialists are being supplied by a private contractor. The objective of the Survey is to identify environmental problems and areas of environmental regulation. It is being performed in accordance with the DOE Environmental Survey Manual. This phase of the Survey involves the review of existing site environmental data, observations of the operations performed at PUEC, and interviews with site personnel. The Survey team developed a Sampling and Analysis Plan to assist in further assessing certain of the environmental problems identified during its on-site activities. The Sampling and Analysis Plan will be executed by Argonne National Laboratory. When completed, the results will be incorporated into the PUEC Environmental Survey Interim Report. The Interim Report will reflect the final determinations of the PUEC Survey. 55 refs., 22 figs., 21 tabs.

  15. New Measures to Safeguard Gas Centrifuge Enrichment Plants

    SciTech Connect (OSTI)

    Whitaker, Jr., James; Garner, James R; Whitaker, Michael; Lockwood, Dunbar; Gilligan, Kimberly V; Younkin, James R; Hooper, David A; Henkel, James J; Krichinsky, Alan M

    2011-01-01

    As Gas Centrifuge Enrichment Plants (GCEPs) increase in separative work unit (SWU) capacity, the current International Atomic Energy Agency (IAEA) model safeguards approach needs to be strengthened. New measures to increase the effectiveness of the safeguards approach are being investigated that will be mutually beneficial to the facility operators and the IAEA. One of the key concepts being studied for application at future GCEPs is embracing joint use equipment for process monitoring of load cells at feed and withdrawal (F/W) stations. A mock F/W system was built at Oak Ridge National Laboratory (ORNL) to generate and collect F/W data from an analogous system. The ORNL system has been used to collect data representing several realistic normal process and off-normal (including diversion) scenarios. Emphasis is placed on the novelty of the analysis of data from the sensors as well as the ability to build information out of raw data, which facilitates a more effective and efficient verification process. This paper will provide a progress report on recent accomplishments and next steps.

  16. Taguchi methods applied to oxygen-enriched diesel engine experiments

    SciTech Connect (OSTI)

    Marr, W.W.; Sekar, R.R.; Cole, R.L.; Marciniak, T.J.; Longman, D.E.

    1992-12-01

    This paper describes a test series conducted on a six-cylinder diesel engine to study the impacts of controlled factors (i.e., oxygen content of the combustion air, water content of the fuel, fuel rate, and fuel-injection timing) on engine emissions using Taguchi methods. Three levels of each factor were used in the tests. Only the main effects of the factors were examined; no attempt was made to analyze the interactions among the factors. It was found that, as in the case of the single-cylinder engine tests, oxygen in the combustion air was very effective in reducing particulate and smoke emissions. Increases in NO{sub x} due to the oxygen enrichment observed in the single-cylinder tests also occurred in the present six-cylinder tests. Water in the emulsified fuel was found to be much less effective in decreasing NO{sub x} emissions for the six-cylinder engine than it was for the single-cylinder engine.

  17. Taguchi methods applied to oxygen-enriched diesel engine experiments

    SciTech Connect (OSTI)

    Marr, W.W.; Sekar, R.R.; Cole, R.L.; Marciniak, T.J. ); Longman, D.E. )

    1992-01-01

    This paper describes a test series conducted on a six-cylinder diesel engine to study the impacts of controlled factors (i.e., oxygen content of the combustion air, water content of the fuel, fuel rate, and fuel-injection timing) on engine emissions using Taguchi methods. Three levels of each factor were used in the tests. Only the main effects of the factors were examined; no attempt was made to analyze the interactions among the factors. It was found that, as in the case of the single-cylinder engine tests, oxygen in the combustion air was very effective in reducing particulate and smoke emissions. Increases in NO[sub x] due to the oxygen enrichment observed in the single-cylinder tests also occurred in the present six-cylinder tests. Water in the emulsified fuel was found to be much less effective in decreasing NO[sub x] emissions for the six-cylinder engine than it was for the single-cylinder engine.

  18. Isotope-enriched protein standards for computational amide I spectroscopy

    SciTech Connect (OSTI)

    Reppert, Mike; Roy, Anish R.; Tokmakoff, Andrei

    2015-03-28

    We present a systematic isotope labeling study of the protein G mutant NuG2b as a step toward the production of reliable, structurally stable, experimental standards for amide I infrared spectroscopic simulations. By introducing isotope enriched amino acids into a minimal growth medium during bacterial expression, we induce uniform labeling of the amide bonds following specific amino acids, avoiding the need for chemical peptide synthesis. We use experimental data to test several common amide I frequency maps and explore the influence of various factors on map performance. Comparison of the predicted absorption frequencies for the four maps tested with empirical assignments to our experimental spectra yields a root-mean-square error of 6-12 cm{sup −1}, with outliers of at least 12 cm{sup −1} in all models. This means that the predictions may be useful for predicting general trends such as changes in hydrogen bonding configuration; however, for finer structural constraints or absolute frequency assignments, the models are unreliable. The results indicate the need for careful testing of existing literature maps and shed light on possible next steps for the development of quantitative spectral maps.

  19. The Need for US Actinide Enrichment Capability (Conference) | SciTech

    Office of Scientific and Technical Information (OSTI)

    Connect The Need for US Actinide Enrichment Capability Citation Details In-Document Search Title: The Need for US Actinide Enrichment Capability Authors: Patton, Bradley D [1] ; Robinson, Sharon M [1] + Show Author Affiliations ORNL Publication Date: 2014-01-01 OSTI Identifier: 1156743

  20. Sensitive Targeted Quantification of ERK Phosphorylation Dynamics and Stoichiometry in Human Cells without Affinity Enrichment

    SciTech Connect (OSTI)

    Shi, Tujin; Gao, Yuqian; Gaffrey, Matthew J.; Nicora, Carrie D.; Fillmore, Thomas L.; Chrisler, William B.; Gritsenko, Marina A.; Wu, Chaochao; He, Jintang; Bloodsworth, Kent J.; Zhao, Rui; Camp, David G.; Liu, Tao; Rodland, Karin D.; Smith, Richard D.; Wiley, H. S.; Qian, Weijun

    2014-12-17

    Mass spectrometry-based targeted quantification is a promising technology for site-specific quantification of posttranslational modifications (PTMs). However, a major constraint of most targeted MS approaches is the limited sensitivity for quantifying low-abundance PTMs, requiring the use of affinity reagents to enrich specific PTMs. Herein, we demonstrate the direct site-specific quantification of ERK phosphorylation isoforms (pT, pY, pTpY) and their relative stoichiometries using a highly sensitive targeted MS approach termed high-pressure, high-resolution separations with intelligent selection and multiplexing (PRISM). PRISM provides effective enrichment of target peptides within a given fraction from complex biological matrix with minimal sample losses, followed by selected reaction monitoring (SRM) quantification. The PRISM-SRM approach enabled direct quantification of ERK phosphorylation in human mammary epithelial cells (HMEC) from as little as 25 µg tryptic peptides from whole cell lysates. Compared to immobilized metal-ion affinity chromatography, PRISM provided >10-fold improvement in signal intensities, presumably due to the better peptide recovery of PRISM for handling small size samples. This approach was applied to quantify ERK phosphorylation dynamics in HMEC treated by different doses of EGF at both the peak activation (10 min) and steady state (2 h). At 10 min, the maximal ERK activation was observed with 0.3 ng/mL dose, whereas the maximal steady state level of ERK activation at 2 h was at 3 ng/ml dose, corresponding to 1200 and 9000 occupied receptors, respectively. At 10 min, the maximally activated pTpY isoform represented ~40% of total ERK, falling to less than 10% at 2 h. The time course and dose-response profiles of individual phosphorylated ERK isoforms indicated that singly phosphorylated pT-ERK never increases significantly, while the increase of pY-ERK paralleled that of pTpY-ERK. This data supports for a processive, rather than distributed, model of ERK phosphorylation. The PRISM-SRM quantification of protein phosphorylation illustrates the potential for simultaneous quantification of multiple PTMs.

  1. A Non-Proliferating Fuel Cycle: No Enrichment, Reprocessing or Accessible Spent Fuel - 12375

    SciTech Connect (OSTI)

    Parker, Frank L.

    2012-07-01

    Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storage sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded emissions), long term care, reduced access to 'dirty' bomb materials, the social and political costs of siting new facilities and the psychological impact of no solution to the nuclear waste problem, were taken into account, the costs would be far lower than those of the present fuel cycle. (authors)

  2. Developing a laser shockwave model for characterizing diffusion bonded interfaces

    SciTech Connect (OSTI)

    Lacy, Jeffrey M. Smith, James A. Rabin, Barry H.

    2015-03-31

    The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) with the goal of reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU in high-power research reactors. The new LEU fuel is a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to support the fuel qualification process, the Laser Shockwave Technique (LST) is being developed to characterize the clad-clad and fuel-clad interface strengths in fresh and irradiated fuel plates. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves to characterize interfaces in nuclear fuel plates. However, because the deposition of laser energy into the containment layer on a specimen's surface is intractably complex, the shock wave energy is inferred from the surface velocity measured on the backside of the fuel plate and the depth of the impression left on the surface by the high pressure plasma pulse created by the shock laser. To help quantify the stresses generated at the interfaces, a finite element method (FEM) model is being utilized. This paper will report on initial efforts to develop and validate the model by comparing numerical and experimental results for back surface velocities and front surface depressions in a single aluminum plate representative of the fuel cladding.

  3. DOE to Remove 200 Metric Tons of Highly Enriched Uranium from...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Will Be Redirected to Naval Reactors, Down-blended or Used for Space Programs WASHINGTON, ... Approximately 20 MT will be reserved for space and research reactors that currently use ...

  4. Robotic Enrichment Processing of Roche 454 Titanium Emlusion PCR at the DOE Joint Genome Institute

    SciTech Connect (OSTI)

    Hamilton, Matthew; Wilson, Steven; Bauer, Diane; Miller, Don; Duffy-Wei, Kecia; Hammon, Nancy; Lucas, Susan; Pollard, Martin; Cheng, Jan-Fang

    2010-05-28

    Enrichment of emulsion PCR product is the most laborious and pipette-intensive step in the 454 Titanium process, posing the biggest obstacle for production-oriented scale up. The Joint Genome Institute has developed a pair of custom-made robots based on the Microlab Star liquid handling deck manufactured by Hamilton to mediate the complexity and ergonomic demands of the 454 enrichment process. The robot includes a custom built centrifuge, magnetic deck positions, as well as heating and cooling elements. At present processing eight emulsion cup samples in a single 2.5 hour run, these robots are capable of processing up to 24 emulsion cup samples. Sample emulsions are broken using the standard 454 breaking process and transferred from a pair of 50ml conical tubes to a single 2ml tube and loaded on the robot. The robot performs the enrichment protocol and produces beads in 2ml tubes ready for counting. The robot follows the Roche 454 enrichment protocol with slight exceptions to the manner in which it resuspends beads via pipette mixing rather than vortexing and a set number of null bead removal washes. The robotic process is broken down in similar discrete steps: First Melt and Neutralization, Enrichment Primer Annealing, Enrichment Bead Incubation, Null Bead Removal, Second Melt and Neutralization and Sequencing Primer Annealing. Data indicating our improvements in enrichment efficiency and total number of bases per run will also be shown.

  5. Houston Pre-Freshman Enrichment Program (Houston PREP). Final report, June 9, 1997--July 25, 1997

    SciTech Connect (OSTI)

    1997-10-01

    The 1997 Houston Pre-Freshman Enrichment Program (PREP) was conducted at the campus of the University of Houston-Downtown from June 9 to July 25, 1997. Program participants were recruited from the Greater Houston Area. All participants were identified as high-achieving students with an interest in learning about the engineering and science professions. The goal of the program was to better prepare our pre-college youth prior to entering college as mathematics, science and engineering majors. The program participants were middle school and high school students from the Aldine, Alief, Channel View, Clear Creek, Cypress-Fairbanks, Fort Bend, Galena Park, Houston, Humble, Katy, Klein, North Forest, Pasadena, Private, and Spring Branch Independent School Districts. Of the 194 students starting the program, 165 students were from economically and socially disadvantage groups under-represented in the engineering and science professions, and 118 of the 194 were women. Our First Year group for 1997 composed of 96% minority and women students. Second and Third Year students combined were 96% minority or women. With financial support from the Center for Computational Sciences and Advanced Distributed Simulation, the Fourth Year Program was added to PREP this year. Twelve students completed the program (83% minority or women).

  6. Student science enrichment training program. Progress report, June 1, 1991--May 31, 1992

    SciTech Connect (OSTI)

    Sandhu, S.S.

    1992-04-21

    Historically Black Colleges and Universities wing of the United States Department of Energy (DOE) provided funds to Claflin College, Orangeburg, S.C. To conduct a student Science Enrichment Training Program for a period of six weeks during 1991 summer. Thirty participants were selected from a pool of applicants, generated by the High School Seniors and Juniors and the Freshmen class of 1990-1991 at Claflin College. The program primarily focused on high ability students, with potential for Science, Mathematics and Engineering Careers. The major objectives of the program were W to increase the pool of well qualified college entering minority students who will elect to go in Physical Sciences and Engineering and (II) to increase the enrollment in Chemistry and Preprofessional-Pre-Med, Pre-Dent, etc.-majors at Claflin College by including the Claflin students to participate in summer academic program. The summer academic program consisted of Chemistry and Computer Science training. The program placed emphasis upon laboratory experience and research. Visits to Scientific and Industrial laboratories were arranged. Guest speakers which were drawn from academia, industry and several federal agencies, addressed the participants on the future role of Science in the industrial growth of United States of America. The guest speakers also acted as role models for the participants. Several videos and films, emphasizing the role of Science in human life, were also screened.

  7. Enhanced Algorithm for Traceability Measurements in UF6 Flow Pipe

    SciTech Connect (OSTI)

    Copinger, Thomas E; March-Leuba, Jose A; Upadhyaya, Belle R

    2007-01-01

    The Blend Down Monitoring System (BDMS) is used to continually assess the mixing and downblending of highly enriched uranium (HEU) with low-enriched uranium (LEU). This is accomplished by measuring the enrichment and the fissile mass flow rate of the UF{sub 6} gas located in each process pipe of the system by inducing the fission of the {sup 235}U contained in the gas. Measurements are taken along this process route to trace the HEU content all the way to the product stream, ensuring that HEU was down blended. A problem associated with the current traceability measuring algorithm is that it does not account for the time-varying background that is introduced to the system by the movement of the shutter located at the HEU leg of the process. The current way of dealing with that problem is to discard the data for periods when the HEU shutter is open (50% of overall data) because it correlates with the same timeframe in which the direct contribution to background from the HEU shutter was seen. The advanced algorithm presented in this paper allows for continuous measurement of traceability (100%) by accurately accounting for the varying background during the shutter-movement cycle. This algorithm utilizes advanced processing techniques that identify and discriminate the different sources of background radiation, instead of grouping them into one background group for the whole measurement cycle. By using this additional information, the traceability measurement statistics can achieve a greater number of values, thus improving the overall usefulness of these measurements in the BDMS. The effectiveness of the new algorithm was determined by modeling it in a simulation and ensuring that it retained its integrity through a large number of runs, including various shutter-failure conditions. Each run was performed with varying amounts of background radiation from each individual source and with varying traceability counts. The simulations documented in this paper prove that the algorithm can stand up to various transients introduced into the system, such as failure of shutter movement.

  8. FACE: Free-Air CO[sub 2] Enrichment for plant research in the field

    SciTech Connect (OSTI)

    Hendrey, G.R.

    1992-08-01

    Research programs concerning the effects of Carbon Dioxide(CO)[sub 2] on cotton plants are described. Biological responses studied include foliage response to CO[sub 2] fluctuations; yield of cotton exposed to CO[sub 2] enrichment; responses of photosynthesis and stomatal conductance to elevated CO[sub 2] in field-grown cotton; cotton leaf and boll temperatures; root response to CO[sub 2] enrichment; and evaluations of cotton response to CO[sub 2] enrichment with canopy reflectance observations.

  9. FACE: Free-Air CO{sub 2} Enrichment for plant research in the field

    SciTech Connect (OSTI)

    Hendrey, G.R.

    1992-08-01

    Research programs concerning the effects of Carbon Dioxide(CO){sub 2} on cotton plants are described. Biological responses studied include foliage response to CO{sub 2} fluctuations; yield of cotton exposed to CO{sub 2} enrichment; responses of photosynthesis and stomatal conductance to elevated CO{sub 2} in field-grown cotton; cotton leaf and boll temperatures; root response to CO{sub 2} enrichment; and evaluations of cotton response to CO{sub 2} enrichment with canopy reflectance observations.

  10. EA-1172: Sale of Surplus Natural and Low Enriched Uranium, Piketon, Ohio

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts for the proposal to sell uranium for subsequent enrichment and fabrication into commercial nuclear power reactor fuel.  The uranium is currently stored...

  11. EA-1255: Project Partnership Transportation of Foreign-Owned Enriched Uranium from the Republic of Georgia

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts for the proposal to transport 5.26 kilograms of enriched uranium-23 5 in the form of nuclear fuel, from the Republic of Georgia to the United Kingdom.

  12. Isotope enrichment by frequency-tripled temperature tuned neodymium laser photolysis of formaldehyde

    DOE Patents [OSTI]

    Marling, John B.

    1977-01-01

    Enrichment of carbon, hydrogen and/or oxygen isotopes by means of isotopically selective photo-predissociation of formaldehyde is achieved by irradiation provided by a frequency-tripled, temperature tuned neodymium laser.

  13. Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2011 Financial Statement Audit

    Energy Savers [EERE]

    Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2011 Financial Statement Audit OAS-FS-13-02 October 2012 September 7, 2012 Mr. Gregory Friedman Inspector General U.S. Department of Energy 1000 Independence Avenue, S.W. Room 5D-039 Washington, DC 20585 Dear Mr. Friedman: We have audited the financial statements of the Department of Energy's (the Department) Uranium Enrichment Decontamination and Decommissioning Fund (D&D Fund) as of and for the year ended September

  14. Jefferson Lab Offers Science Enrichment Program for 5th, 6th & 8th Grade

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Teachers | Jefferson Lab amp; 8th Grade Teachers Jefferson Lab Offers Science Enrichment Program for 5th, 6th & 8th Grade Teachers NEWPORT NEWS, Va., UPDATED August 4, 2010 - The U.S. Department of Energy's Jefferson Lab is currently accepting applications for its science enrichment program for fifth-, sixth- and eighth-grade teachers of science. The program, designed to increase teachers' knowledge of the physical sciences and strengthen their teaching skills, runs from September 2010

  15. Jefferson Lab Offers Science Enrichment Program for 5th, 6th & 8th Grade

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Teachers | Jefferson Lab Teachers Get Their Science On - One hundred sixty-two elementary and middle-school teachers interested in learning new and innovative methods for teaching the physical sciences attended the Annual Region II Teacher Night held April 20 at Jefferson Lab. Fifty-four teachers who participated in enrichment programs at JLab for teachers of science presented the activities and demonstrations. Jefferson Lab Offers Science Enrichment Program for 5th, 6th & 8th Grade

  16. Jefferson Lab Offers Science Enrichment Program for 5th, 6th & 8th Grade

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Teachers; Registration Deadline is Sept. 13 | Jefferson Lab 3 Jefferson Lab Offers Science Enrichment Program for 5th, 6th & 8th Grade Teachers; Registration Deadline is Sept. 13 NEWPORT NEWS, Va., Aug. 28, 2013 -- The U.S. Department of Energy's Jefferson Lab is currently accepting applications for its science enrichment program for fifth-, sixth- and eighth-grade teachers of science. The after-school program is designed to increase teachers' knowledge of the physical sciences and to

  17. Toxic Substances Control Act Uranium Enrichment Federal Facilities Compliance Agreement, February 20, 1992 Summary

    Office of Environmental Management (EM)

    Toxic Substance Control Act Uranium Enrichment Federal Facilities Compliance Agreement (TSCA-UE- FFCA), February 20, 1992 State Kentucky Agreement Type Compliance Agreement Legal Driver(s) TSCA Scope Summary Establishes responsibilities and commitments for bringing DOE's former and active Uranium Enrichment Plants in Paducah, Portsmouth, and Oak Ridge into compliance with TSCA and PCB Regulations Parties DOE; U.S. EPA Date 2/20/1992 SCOPE * Establish a plan and the responsibilities and

  18. Advanced Neutron Source enrichment study -- Volume 1: Main report. Final report, Revision 12/94

    SciTech Connect (OSTI)

    Bari, R.A.; Ludewig, H.; Weeks, J.

    1994-12-31

    A study has been performed of the impact on performance of using low enriched uranium (20% {sup 235}U) or medium enriched uranium (35% {sup 235}U) as an alternative fuel for the Advanced Neutron Source, which is currently designed to use uranium enriched to 93% {sup 235}U. Higher fuel densities and larger volume cores were evaluated at the lower enrichments in terms of impact on neutron flux, safety, safeguards, technical feasibility, and cost. The feasibility of fabricating uranium silicide fuel at increasing material density was specifically addressed by a panel of international experts on research reactor fuels. The most viable alternative designs for the reactor at lower enrichments were identified and discussed. Several sensitivity analyses were performed to gain an understanding of the performance of the reactor at parametric values of power, fuel density, core volume, and enrichment that were interpolations between the boundary values imposed on the study or extrapolations from known technology. Volume 2 of this report contains 26 appendices containing results, meeting minutes, and fuel panel presentations.

  19. Statistical Hot Channel Analysis for the NBSR

    SciTech Connect (OSTI)

    Cuadra A.; Baek J.

    2014-05-27

    A statistical analysis of thermal limits has been carried out for the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The objective of this analysis was to update the uncertainties of the hot channel factors with respect to previous analysis for both high-enriched uranium (HEU) and low-enriched uranium (LEU) fuels. Although uncertainties in key parameters which enter into the analysis are not yet known for the LEU core, the current analysis uses reasonable approximations instead of conservative estimates based on HEU values. Cumulative distribution functions (CDFs) were obtained for critical heat flux ratio (CHFR), and onset of flow instability ratio (OFIR). As was done previously, the Sudo-Kaminaga correlation was used for CHF and the Saha-Zuber correlation was used for OFI. Results were obtained for probability levels of 90%, 95%, and 99.9%. As an example of the analysis, the results for both the existing reactor with HEU fuel and the LEU core show that CHFR would have to be above 1.39 to assure with 95% probability that there is no CHF. For the OFIR, the results show that the ratio should be above 1.40 to assure with a 95% probability that OFI is not reached.

  20. Development of a New Multiplying Assembly for Research, Validation, Evaluation, and Learning

    SciTech Connect (OSTI)

    David L. Chichester

    2012-10-01

    A new multiplying test assembly is under development at Idaho National Laboratory (INL) to support research, validation, evaluation, and learning. The item is comprised of two stacked highly-enriched uranium (HEU) cylinders each 11.4 cm in diameter and having a combined height of 8.4 cm. The combined mass is 14.4 kg of HEU. Calculations for the bare configuration of the assembly indicate a multiplication level of >2.5 (keff = 0.62). Reflected configurations of the assembly, using either polyethylene or tungsten, are possible and have the capability of raising its multiplication level to approximately 8. This paper will describe the MCNP calculations performed to assess the assembly's multiplication level under different conditions and describe the resource available at INL to support visiting researchers in their use of the material. We will also describe some preliminary calculations and test activities using the assembly to study neutron multiplicity.

  1. Protein scaffolds for selective enrichment of metal ions

    DOE Patents [OSTI]

    He, Chuan; Zhou, Lu; Bosscher, Michael

    2016-02-09

    Polypeptides comprising high affinity for the uranyl ion are provided. Methods for binding uranyl using such proteins are likewise provided and can be used, for example, in methods for uranium purification or removal.

  2. Use of bromodeoxyuridine immunocapture to identify psychrotolerant phenanthrene-degrading bacteria in phenanthrene-enriched polluted Baltic Sea sediments

    SciTech Connect (OSTI)

    Edlund, A.; Jansson, J.

    2008-05-01

    The aim of this study was to enrich and identify psychrotolerant phenanthrenedegrading bacteria from polluted Baltic Sea sediments. Polyaromatic hydrocarbon (PAH)-contaminated sediments were spiked with phenanthrene and incubated for 2 months in the presence of bromodeoxyuridine that is incorporated into the DNA of replicating cells. The bromodeoxyuridine-incorporated DNA was extracted by immunocapture and analyzed by terminal-restriction fragment length polymorphism and 16S rRNA gene cloning and sequencing to identify bacterial populations that were growing. In addition, degradation genes were quantified in the bromodeoxyuridine-incorporated DNA by real-time PCR. Phenanthrene concentrations decreased after 2 months of incubation in the phenanthrene-enriched sediments and this reduction correlated to increases in copy numbers of xylE and phnAc dioxygenase genes. Representatives of Exiguobacterium, Schewanella,Methylomonas, Pseudomonas, Bacteroides and an uncultured Deltaproteobacterium and a Gammaproteobacterium dominated the growing community in the phenanthrene spiked sediments. Isolates that were closely related to three of these bacteria (two pseudomonads and an Exiguobacterium sp.) could reduce phenanthrene concentrations in pure cultures and they all harbored phnAc dioxygenase genes. These results confirm that this combination of culture-based and molecular approaches was useful for identification of actively growing bacterial species with a high potential for phenanthrene degradation.

  3. ELEMENTAL ABUNDANCES AND THEIR IMPLICATIONS FOR THE CHEMICAL ENRICHMENT OF THE BOOeTES I ULTRAFAINT GALAXY

    SciTech Connect (OSTI)

    Gilmore, Gerard; Norris, John E.; Yong, David; Monaco, Lorenzo; Wyse, Rosemary F. G.; Geisler, D. E-mail: jen@mso.anu.edu.au E-mail: lmonaco@eso.org E-mail: dgeisler@astro-udec.cl

    2013-01-20

    We present a double-blind analysis of high-dispersion spectra of seven red giant members of the Booetes I ultrafaint dwarf spheroidal galaxy, complemented with re-analysis of a similar spectrum of an eighth-member star. The stars cover [Fe/H] from -3.7 to -1.9 and include a CEMP-no star with [Fe/H] = -3.33. We conclude from our chemical abundance data that Booetes I has evolved as a self-enriching star-forming system, from essentially primordial initial abundances. This allows us uniquely to investigate the place of CEMP-no stars in a chemically evolving system, in addition to limiting the timescale of star formation. The elemental abundances are formally consistent with a halo-like distribution, with enhanced mean [{alpha}/Fe] and small scatter about the mean. This is in accord with the high-mass stellar initial mass function in this low-stellar-density, low-metallicity system being indistinguishable from the present-day solar neighborhood value. There is a non-significant hint of a decline in [{alpha}/Fe] with [Fe/H]; together with the low scatter, this requires low star formation rates, allowing time for supernova ejecta to be mixed over the large spatial scales of interest. One star has very high [Ti/Fe], but we do not confirm a previously published high value of [Mg/Fe] for another star. We discuss the existence of CEMP-no stars, and the absence of any stars with lower CEMP-no enhancements at higher [Fe/H], a situation that is consistent with knowledge of CEMP-no stars in the Galactic field. We show that this observation requires there be two enrichment paths at very low metallicities: CEMP-no and 'carbon-normal'.

  4. HEU Minimization and the Reliable Supply of Medical Isotopes...

    National Nuclear Security Administration (NNSA)

    This Site Budget IG Web Policy Privacy No Fear Act Accessibility FOIA Sitemap Federal Government The White House DOE.gov USA.gov Jobs Apply for Our Jobs Our Jobs Working at NNSA...

  5. Amended ROD for HEU EIS (8-22-11).pdf

    National Nuclear Security Administration (NNSA)

  6. NNSA Recognizes Industry Partners for Efforts to Eliminate HEU...

    National Nuclear Security Administration (NNSA)

    ... application of nuclear science. NNSA maintains and enhances the safety, security, reliability and performance of the U.S. nuclear weapons stockpile without nuclear testing; works ...

  7. Last HEU Removed from Switzerland under NNSA Collaboration |...

    National Nuclear Security Administration (NNSA)

    application of nuclear science. NNSA maintains and enhances the safety, security, reliability and performance of the U.S. nuclear weapons stockpile without nuclear testing; works ...

  8. NNSA Partnership Successfully Removes All Remaining HEU from...

    National Nuclear Security Administration (NNSA)

    ... application of nuclear science. NNSA maintains and enhances the safety, security, reliability and performance of the U.S. nuclear weapons stockpile without nuclear testing; works ...

  9. U.S. HEU Disposition Program | National Nuclear Security Administratio...

    National Nuclear Security Administration (NNSA)

    People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy ... In 1996, the Department of Energy (DOE) announced plans to reduce the proliferation threat ...

  10. Modeling of UF{sub 6} enrichment with gas centrifuges for nuclear safeguards activities

    SciTech Connect (OSTI)

    Mercurio, G.; Peerani, P.; Richir, P.; Janssens, W.; Eklund, G.

    2012-09-26

    The physical modeling of uranium isotopes ({sup 235}U, {sup 238}U) separation process by centrifugation of is a key aspect for predicting the nuclear fuel enrichment plant performances under surveillance by the Nuclear Safeguards Authorities. In this paper are illustrated some aspects of the modeling of fast centrifuges for UF{sub 6} gas enrichment and of a typical cascade enrichment plant with the Theoretical Centrifuge and Cascade Simulator (TCCS). The background theory for reproducing the flow field characteristics of a centrifuge is derived from the work of Cohen where the separation parameters are calculated using the solution of a differential enrichment equation. In our case we chose to solve the hydrodynamic equations for the motion of a compressible fluid in a centrifugal field using the Berman - Olander vertical velocity radial distribution and the solution was obtained using the Matlab software tool. The importance of a correct estimation of the centrifuge separation parameters at different flow regimes, lies in the possibility to estimate in a reliable way the U enrichment plant performances, once the separation external parameters are set (feed flow rate and feed, product and tails assays). Using the separation parameters of a single centrifuge allow to determine the performances of an entire cascade and, for this purpose; the software Simulink was used. The outputs of the calculation are the concentrations (assays) and the flow rates of the enriched (product) and depleted (tails) gas mixture. These models represent a valid additional tool, in order to verify the compliance of the U enrichment plant operator declarations with the 'on site' inspectors' measurements.

  11. Houston prefreshman enrichment program (Houston PREP). Final report, June 10, 1996--August 1, 1996

    SciTech Connect (OSTI)

    1996-10-01

    The 1996 Houston Pre-freshman Enrichment Program (PREP) was conducted on the campus of the University of Houston-Downtown from June 10 to August 1, 1996. Program Participants were recruited from the Greater Houston area. All participants were identified as high achieving students with an interest in learning about the engineering and science professions. The goal of the program was to better prepare our pre-college youth prior to entering college as mathematics, science and engineering majors. The program participants were middle school and high school students from the Aldine, Alief, Channel View, Crockett, Cypress-Fairbanks, Fort Bend, Galena Park, Houston, Humble, Katy, Klein, North Forest, Pasadena, Private, and Spring Branch Independent School Districts. Of the 197 students starting the program, 170 completed, 142 students were from economically and socially disadvantage groups underrepresented in the engineering and science professions, and 121 of the 197 were female. Our First Year group for 1996 composed of 96% minority and women students. Our Second and Third Year students were 100% and 93.75% minority or women respectively. This gave an overall minority and female population of 93.75%. This year, special efforts were again made to recruit students from minority groups, which caused a significant increase in qualified applicants. However, due to space limitations, 140 applicants were rejected. Investigative and discovery learning were key elements of PREP. The academic components of the program included Algebraic Structures, Engineering, Introduction to Computer Science, Introduction to Physics, Logic and Its Application to Mathematics, Probability and Statistics, Problem Solving Seminar using computers and PLATO software, SAT Preparatory Seminars, and Technical Writing.

  12. Cost comparisons among free-air CO{sub 2} enrichment, open-top chamber, and sunlit controlled-environment chamber methods of CO{sub 2} exposure

    SciTech Connect (OSTI)

    Kimball, B.A.

    1992-12-31

    Several techniques for exposing plants and ecosystems to elevated CO{sub 2} were presented and the experimental characteristics of each were discussed in detail. However, one important factor which was not addressed is cost, which can easily dictate use of less than ideal methods. The purpose of this chapter is to evaluate and compare the approximate costs of the free-air CO{sub 2} enrichment (FACE) method with two other popular techniques - open-top chambers (OTC) and sunlit controlled-environment chambers (Soil, Plant, Atmosphere Research chambers, SPAR). The primary comparison among the SPAR, OTC, and FACE techniques is a consideration of the relative costs incurred in the production of high-CO{sub 2}-grown plants. However, measurements of plant response must also be made, which requires scientific personnel and instruments, so the plant production costs will also be compared to the overall costs of conducting CO{sub 2}-enrichment experiments. 2 refs., 1 fig., 1 tab.

  13. Simulation studies of diesel engine performance with oxygen enriched air and water emulsified fuels

    SciTech Connect (OSTI)

    Assanis, D.N.; Baker, D. ); Sekar, R.R.; Siambekos, C.T.; Cole, R.L.; Marciniak, T.J. )

    1990-01-01

    A computer simulation code of a turbocharged, turbocompound diesel engine was modified to study the effects of using oxygen-enriched combustion air and water-emulsified diesel fuels. Oxygen levels of 21 percent to 40 percent by volume in the combustion air were studied. Water content in the fuel was varied from 0 percent to 50 percent mass. Simulation studies and a review and analysis of previous work in this area led to the following conclusions about expected engine performance and emissions: the power density of the engine is significantly increased by oxygen enrichment. Ignition delay and particulate emissions are reduced. Combustion temperatures and No{sub x} emissions are increased with oxygen enrichment but could be brought back to the base levels by introducing water in the fuel. The peak cylinder pressure which increases with the power output level might result in mechanical problems with engine components. Oxygen enrichment also provides an opportunity to use cheaper fuel such as No. 6 diesel fuel. Overall, the adverse effects of oxygen enrichment could be countered by the addition of water and it appears that an optimum combination of water content, oxygen level, and base diesel fuel quality may exist. This could yield improved performance and emissions characteristics compared to a state-of-the-art diesel engine. 9 refs., 8 figs.

  14. Metal affinity enrichment increases the range and depth of proteome identification for extracellular microbial proteins

    SciTech Connect (OSTI)

    Wheeler, Korin; Erickson, Brian K; Mueller, Ryan; Singer, Steven; Verberkmoes, Nathan C; Hwang, Mona; Thelen, Michael P.; Hettich, Robert {Bob} L

    2012-01-01

    Many key proteins, such as those involved in cellular signaling or transcription, are difficult to measure in microbial proteomic experiments due to the interfering presence of more abundant, dominant proteins. In an effort to enhance the identification of previously undetected proteins, as well as provide a methodology for selective enrichment, we evaluated and optimized immobilized metal affinity chromatography (IMAC) coupled with mass spectrometric characterization of extracellular proteins from an extremophilic microbial community. Seven different metals were tested for IMAC enrichment. The combined results added 20% greater proteomic depth to the extracellular proteome. Although this IMAC enrichment could not be conducted at the physiological pH of the environmental system, this approach did yield a reproducible and specific enrichment of groups of proteins with functions potentially vital to the community, thereby providing a more extensive biochemical characterization. Notably, 40 unknown proteins previously annotated as hypothetical were enriched and identified for the first time. Examples of identified proteins includes a predicted TonB signal sensing protein homologous to other known TonB proteins and a protein with a COXG domain previously identified in many chemolithoautotrophic microbes as having a function in the oxidation of CO.

  15. Axi-symmetrical flow reactor for [sup 196]Hg photochemical enrichment

    DOE Patents [OSTI]

    Grossman, M.W.

    1991-04-30

    The present invention is directed to an improved photochemical reactor useful for the isotopic enrichment of a predetermined isotope of mercury, especially, [sup 196]Hg. Specifically, two axi-symmetrical flow reactors were constructed according to the teachings of the present invention. These reactors improve the mixing of the reactants during the photochemical enrichment process, affording higher yields of the desired [sup 196]Hg product. Measurements of the variation of yield (Y) and enrichment factor (E) along the flow axis of these reactors indicates very substantial improvement in process uniformity compared to previously used photochemical reactor systems. In one preferred embodiment of the present invention, the photoreactor system was built such that the reactor chamber was removable from the system without disturbing the location of either the photochemical lamp or the filter employed therewith. 10 figures.

  16. Axi-symmetrical flow reactor for .sup.196 Hg photochemical enrichment

    DOE Patents [OSTI]

    Grossman, Mark W.

    1991-01-01

    The present invention is directed to an improved photochemical reactor useful for the isotopic enrichment of a predetermined isotope of mercury, especially, .sup.196 Hg. Specifically, two axi-symmetrical flow reactors were constructed according to the teachings of the present invention. These reactors improve the mixing of the reactants during the photochemical enrichment process, affording higher yields of the desired .sup.196 Hg product. Measurements of the variation of yield (Y) and enrichment factor (E) along the flow axis of these reactors indicates very substantial improvement in process uniformity compared to previously used photochemical reactor systems. In one preferred embodiment of the present invention, the photoreactor system was built such that the reactor chamber was removable from the system without disturbing the location of either the photochemical lamp or the filter employed therewith.

  17. Application of atomic vapor laser isotope separation to the enrichment of mercury

    SciTech Connect (OSTI)

    Crane, J.K.; Erbert, G.V.; Paisner, J.A.; Chen, H.L.; Chiba, Z.; Beeler, R.G.; Combs, R.; Mostek, S.D.

    1986-09-01

    Workers at GTE/Sylvania have shown that the efficiency of fluorescent lighting may be markedly improved using mercury that has been enriched in the /sup 196/Hg isotope. A 5% improvement in the efficiency of fluorescent lighting in the United States could provide a savings of approx. 1 billion dollars in the corresponding reduction of electrical power consumption. We will discuss the results of recent work done at our laboratory to develop a process for enriching mercury. The discussion will center around the results of spectroscopic measurements of excited state lifetimes, photoionization cross sections and isotope shifts. In addition, we will discuss the mercury separator and supporting laser mesurements of the flow properties of mercury vapor. We will describe the laser system which will provide the photoionization and finally discuss the economic details of producing enriched mercury at a cost that would be attractive to the lighting industry.

  18. Process for recovering evolved hydrogen enriched with at least one heavy hydrogen isotope

    DOE Patents [OSTI]

    Tanaka, John; Reilly, Jr., James J.

    1978-01-01

    This invention relates to a separation means and method for enriching a hydrogen atmosphere with at least one heavy hydrogen isotope by using a solid titaniun alloy hydride. To this end, the titanium alloy hydride containing at least one metal selected from the group consisting of vanadium, chromium, manganese, molybdenum, iron, cobalt and nickel is contacted with a circulating gaseous flow of hydrogen containing at least one heavy hydrogen isotope at a temperature in the range of -20.degree. to +40.degree. C and at a pressure above the dissociation pressure of the hydrided alloy selectively to concentrate at least one of the isotopes of hydrogen in the hydrided metal alloy. The contacting is continued until equilibrium is reached, and then the gaseous flow is isolated while the temperature and pressure of the enriched hydride remain undisturbed selectively to isolate the hydride. Thereafter, the enriched hydrogen is selectively recovered in accordance with the separation factor (S.F.) of the alloy hydride employed.

  19. Characteristics of enriched cultures for bio-huff-`n`-puff tests at Jilin oil field

    SciTech Connect (OSTI)

    Xiu-Yuan Wang; Gang Dai; Yan-Fen Xue; Shu-Hua Xie

    1995-12-31

    Three enriched cultures (48, 15a, and 26a), selected from more than 80 soil and water samples, could grow anaerobically in the presence of crude oil at 30{degrees}C and could ferment molasses to gases and organic acids. Oil recovery by culture 48 in the laboratory model experiment was enhanced by 25.2% over the original reserves and by 53.7% over the residual reserves. Enriched culture 48 was composed of at least 4 species belonging to the genera Eubacterium, Fusobacterium, and Bacteroides. This enriched culture was used as inoculum for MEOR field trials at Jilin oil field with satisfactory results. The importance of the role of these isolates in EOR was confirmed by their presence and behavior in the fluids produced from the microbiologically treated reservoir.

  20. Synchrotron based infrared imaging and spectroscopy via focal plane array on live fibroblasts in D2O enriched medium

    SciTech Connect (OSTI)

    Quaroni, Luca; Zlateva, Theodora; Sarafimov, Blagoj; Kreuzer, Helen W.; Wehbe, Katia; Hegg, Eric L.; Cinque, Gianfelice

    2014-03-26

    We tested the viability of using synchrotron based infrared imaging to study biochemical processes inside living cells. As a model system, we studied fibroblast cells exposed to a medium highly enriched with D2O. We could show that the experimental technique allows us to reproduce at the cellular level measurements that are normally performed on purified biological molecules. We can obtain information about lipid conformation and distribution, kinetics of hydrogen/deuterium exchange, and the formation of concentration gradients of H and O isotopes in water that are associated with cell metabolism. The implementation of the full field technique in a sequential imaging format gives a description of cellular biochemistry and biophysics that contains both spatial and temporal information.

  1. Critical experiments on an enriched uranium solution system containing periodically distributed strong thermal neutron absorbers

    SciTech Connect (OSTI)

    Rothe, R.E.

    1996-09-30

    A series of 62 critical and critical approach experiments were performed to evaluate a possible novel means of storing large volumes of fissile solution in a critically safe configuration. This study is intended to increase safety and economy through use of such a system in commercial plants which handle fissionable materials in liquid form. The fissile solution`s concentration may equal or slightly exceed the minimum-critical-volume concentration; and experiments were performed for high-enriched uranium solution. Results should be generally applicable in a wide variety of plant situations. The method is called the `Poisoned Tube Tank` because strong neutron absorbers (neutron poisons) are placed inside periodically spaced stainless steel tubes which separate absorber material from solution, keeping the former free of contamination. Eight absorbers are investigated. Both square and triangular pitched lattice patterns are studied. Ancillary topics which closely model typical plant situations are also reported. They include the effect of removing small bundles of absorbers as might occur during inspections in a production plant. Not taking the tank out of service for these inspections would be an economic advantage. Another ancillary topic studies the effect of the presence of a significant volume of unpoisoned solution close to the Poisoned Tube Tank on the critical height. A summary of the experimental findings is that boron compounds were excellent absorbers, as expected. This was true for granular materials such as Gerstley Borate and Borax; but it was also true for the flexible solid composed of boron carbide and rubber, even though only thin sheets were used. Experiments with small bundles of absorbers intentionally removed reveal that quite reasonable tanks could be constructed that would allow a few tubes at a time to be removed from the tank for inspection without removing the tank from production service.

  2. MO-E-18C-06: Enriching Medical Physics Education By Visualizing The Invisible

    SciTech Connect (OSTI)

    Sprawls, P

    2014-06-15

    Purpose: To enhance the understanding of medical physics concepts and develop higher levels of learning relating to invisible physics phenomena such as radiation. To provide medical physics educators in all countries of the world with understanding of knowledge structures in the human brain, the different levels of learning, and the types of knowledge required for higher level functions such as problem solving, creative innovations, and applied clinical applications. To provide medical physics educators with an open access resource (tool) that they can use in their teaching activities to enrich and elevate the level of learning for their students, residents, etc. with respect to the invisible realm of medical physics. Methods: An experienced clinical medical physicist and educator has created and provided with open access three complementary web-based resources to achieve the purposes described above. One is a module focusing on the medical physics learning process with respect to mental knowledge structures, how they relate to outcomes and applications, and learning activities that are required to develop the required knowledge structures. The second is an extensive set of visuals that educators can use in their activities (classes, small group discussions, etc.) to visualize the invisible. The third is an interactive online simulation where learners can adjust factors and visually observe changes in x-radiation.These resources are available online at www.BLINDED FOR REVIEW . Results: Medical physics education, especially for non-physicists, is becoming much more interesting and useful especially with respect to invisible radiation. The global impact is that medical imaging professionals can be more effective in optimizing x-ray imaging procedures and risk management when they have knowledge levels that enhance problem solving, innovation, and creativity. Conclusion: Medical physics educators in all institutions can be much more effective and efficient in the sharing of their knowledge and experience when enhanced with high-quality visuals.

  3. Status Report on the Passive Neutron Enrichment Meter (PNEM) for UF6 Cylinder Assay

    SciTech Connect (OSTI)

    Miller, Karen A.; Swinhoe, Martyn T.; Menlove, Howard O.; Marlow, Johnna B.

    2012-05-02

    The Passive Neutron Enrichment Meter (PNEM) is a nondestructive assay (NDA) system being developed at Los Alamos National Laboratory (LANL). It was designed to determine {sup 235}U mass and enrichment of uranium hexafluoride (UF{sub 6}) in product, feed, and tails cylinders (i.e., 30B and 48Y cylinders). These cylinders are found in the nuclear fuel cycle at uranium conversion, enrichment, and fuel fabrication facilities. The PNEM is a {sup 3}He-based neutron detection system that consists of two briefcase-sized detector pods. A photograph of the system during characterization at LANL is shown in Fig. 1. Several signatures are currently being studied to determine the most effective measurement and data reduction technique for unfolding {sup 235}U mass and enrichment. The system collects total neutron and coincidence data for both bare and cadmium-covered detector pods. The measurement concept grew out of the success of the Uranium Cylinder Assay System (UCAS), which is an operator system at Rokkasho Enrichment Plant (REP) that uses total neutron counting to determine {sup 235}U mass in UF{sub 6} cylinders. The PNEM system was designed with higher efficiency than the UCAS in order to add coincidence counting functionality for the enrichment determination. A photograph of the UCAS with a 48Y cylinder at REP is shown in Fig. 2, and the calibration measurement data for 30B product and 48Y feed and tails cylinders is shown in Fig. 3. The data was collected in a low-background environment, meaning there is very little scatter in the data. The PNEM measurement concept was first presented at the 2010 Institute of Nuclear Materials Management (INMM) Annual Meeting. The physics design and uncertainty analysis were presented at the 2010 International Atomic Energy Agency (IAEA) Safeguards Symposium, and the mechanical and electrical designs and characterization measurements were published in the ESARDA Bulletin in 2011.

  4. COSMOLOGICAL SIMULATIONS OF INTERGALACTIC MEDIUM EVOLUTION. I. TEST OF THE SUBGRID CHEMICAL ENRICHMENT MODEL

    SciTech Connect (OSTI)

    Côté, Benoit; Martel, Hugo; Drissen, Laurent

    2013-11-10

    We present a one-zone galactic chemical enrichment model that takes into account the contribution of stellar winds from massive stars under the effect of rotation, Type II supernovae, hypernovae, stellar winds from low- and intermediate-mass stars, and Type Ia supernovae. This enrichment model will be implemented in a galactic model designed to be used as a subgrid treatment for galaxy evolution and outflow generation in large-scale cosmological simulations, in order to study the evolution of the intergalactic medium. We test our enrichment prescription by comparing its predictions with the metallicity distribution function and the abundance patterns of 14 chemical elements observed in the Milky Way stars. To do so, we combine the effect of many stellar populations created from the star formation history of the Galaxy in the solar neighborhood. For each stellar population, we keep track of its specific mass, initial metallicity, and age. We follow the time evolution of every population in order to respect the time delay between the various stellar phases. Our model is able to reproduce the observed abundances of C, O, Na, Mg, Al, S, and Ca. For Si, Cr, Mn, Ni, Cu, and Zn, the fits are still reasonable, but improvements are needed. We marginally reproduce the nitrogen abundance in very low metallicity stars. Overall, our results are consistent with the predicted abundance ratios seen in previous studies of the enrichment history of the Milky Way. We have demonstrated that our semi-analytic one-zone model, which cannot deal with spatial information such as the metallicity gradient, can nevertheless successfully reproduce the global Galactic enrichment evolution obtained by more complex models, at a fraction of the computational cost. This model is therefore suitable for a subgrid treatment of chemical enrichment in large-scale cosmological simulations.

  5. The Office of Environmental Management Uranium Enrichment D&D | Department

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    of Energy Uranium Enrichment D&D The Office of Environmental Management Uranium Enrichment D&D PDF icon Microsoft Word - B996F741.doc More Documents & Publications Microsoft Word - PSRP Updates 6-25-10_v2 Above on the left is K-25, at Oak Ridge before and after the 844,000 sq-ft demolition. In addition, on the right: K Cooling Tower at Savannah River Site demolition. Deactivation & Decommissioning (D&D) Program Map 2013 Annual Workforce Analysis and Staffing Plan Report -

  6. Deuterium enrichment by selective photo-induced dissociation of an organic carbonyl compound

    DOE Patents [OSTI]

    Marling, John B.

    1981-01-01

    A method for producing a deuterium enriched material by photoinduced dissociation which uses as the working material a gas phase photolytically dissociable organic carbonyl compound containing at least one hydrogen atom bonded to an atom which is adjacent to a carbonyl group and consisting of molecules wherein said hydrogen atom is present as deuterium and molecules wherein said hydrogen atom is present as another isotope of hydrogen. The organic carbonyl compound is subjected to intense infrared radiation at a preselected wavelength to selectively excite and thereby induce dissociation of the deuterium containing species to yield a deuterium enriched stable molecular product. Undissociated carbonyl compound, depleted in deuterium, is preferably redeuterated for reuse.

  7. Active-Interrogation Measurements of Fast Neutrons from Induced Fission in Low-Enriched Uranium

    SciTech Connect (OSTI)

    J. L. Dolan; M. J. Marcath; M. Flaska; S. A. Pozzi; D. L. Chichester; A. Tomanin; P. Peerani

    2014-02-01

    A detection system was designed with MCNPX-PoliMi to measure induced-fission neutrons from U-235 and U-238 using active interrogation. Measurements were then performed with this system at the Joint Research Centre (JRC) in Ispra, Italy on low-enriched uranium samples. Liquid scintillators measured induced fission neutron to characterize the samples in terms of their uranium mass and enrichment. Results are presented to investigate and support the use of organic liquid scintillators with active interrogation techniques to characterize uranium containing materials.

  8. Safeguards at Gas Centrifuge Enrichment Plants: Why is Iran a Threat?

    Office of Scientific and Technical Information (OSTI)

    (Technical Report) | SciTech Connect Technical Report: Safeguards at Gas Centrifuge Enrichment Plants: Why is Iran a Threat? Citation Details In-Document Search Title: Safeguards at Gas Centrifuge Enrichment Plants: Why is Iran a Threat? Authors: Boyer, Brian D. [1] + Show Author Affiliations Los Alamos National Laboratory [Los Alamos National Laboratory Publication Date: 2014-06-25 OSTI Identifier: 1136096 Report Number(s): LA-UR-14-24695 DOE Contract Number: AC52-06NA25396 Resource Type:

  9. Batch methods for enriching trace impurities in hydrogen gas for their further analysis

    DOE Patents [OSTI]

    Ahmed, Shabbir; Lee, Sheldon H.D.; Kumar, Romesh; Papdias, Dionissios D.

    2014-07-15

    Provided herein are batch methods and devices for enriching trace quantities of impurities in gaseous mixtures, such as hydrogen fuel. The methods and devices rely on concentrating impurities using hydrogen transport membranes wherein the time period for concentrating the sample is calculated on the basis of optimized membrane characteristics, comprising its thickness and permeance, with optimization of temperature, and wherein the enrichment of trace impurities is proportional to the pressure ratio P.sub.hi/P.sub.lo and the volume ratio V.sub.1/V.sub.2, with following detection of the impurities using commonly-available detection methods.

  10. THE COS/UVES ABSORPTION SURVEY OF THE MAGELLANIC STREAM. II. EVIDENCE FOR A COMPLEX ENRICHMENT HISTORY OF THE STREAM FROM THE FAIRALL 9 SIGHTLINE

    SciTech Connect (OSTI)

    Richter, Philipp; Fechner, Cora [Institut fuer Physik und Astronomie, Universitaet Potsdam, Haus 28, Karl-Liebknecht-Str. 24/25, D-14476 Golm (Potsdam) (Germany); Fox, Andrew J. [Space Telescope Science Institute, Baltimore, MD 21218 (United States); Wakker, Bart P. [Department of Astronomy, University of Wisconsin-Madison, 475 North Charter Street, Madison, WI 53706 (United States); Lehner, Nicolas; Howk, J. Christopher [Department of Physics, University of Notre Dame, 225 Nieuwland Science Hall, Notre Dame, IN 46556 (United States); Bland-Hawthorn, Joss [Institute of Astronomy, School of Physics, University of Sydney, NSW 2006 (Australia); Ben Bekhti, Nadya [Argelander-Institut fuer Astronomie, Universitaet Bonn, Auf dem Huegel 71, D-53121 Bonn (Germany)

    2013-08-01

    We present a multi-wavelength study of the Magellanic Stream (MS), a massive gaseous structure in the Local Group that is believed to represent material stripped from the Magellanic Clouds. We use ultraviolet, optical and radio data obtained with HST/COS, VLT/UVES, FUSE, GASS, and ATCA to study metal abundances and physical conditions in the Stream toward the quasar Fairall 9. Line absorption in the MS from a large number of metal ions and from molecular hydrogen is detected in up to seven absorption components, indicating the presence of multi-phase gas. From the analysis of unsaturated S II absorption, in combination with a detailed photoionization model, we obtain a surprisingly high {alpha} abundance in the Stream toward Fairall 9 of [S/H] = -0.30 {+-} 0.04 (0.50 solar). This value is five times higher than what is found along other MS sightlines based on similar COS/UVES data sets. In contrast, the measured nitrogen abundance is found to be substantially lower ([N/H] = -1.15 {+-} 0.06), implying a very low [N/{alpha}] ratio of -0.85 dex. The substantial differences in the chemical composition of MS toward Fairall 9 compared to other sightlines point toward a complex enrichment history of the Stream. We favor a scenario, in which the gas toward Fairall 9 was locally enriched with {alpha} elements by massive stars and then was separated from the Magellanic Clouds before the delayed nitrogen enrichment from intermediate-mass stars could set in. Our results support (but do not require) the idea that there is a metal-enriched filament in the Stream toward Fairall 9 that originates in the LMC.

  11. Neutron spectrometry for UF6 enrichment verification in storage cylinders

    SciTech Connect (OSTI)

    Mengesha, Wondwosen; Kiff, Scott D.

    2015-01-29

    Verification of declared UF6 enrichment and mass in storage cylinders is of great interest in nuclear material nonproliferation. Nondestructive assay (NDA) techniques are commonly used for safeguards inspections to ensure accountancy of declared nuclear materials. Common NDA techniques used include gamma-ray spectrometry and both passive and active neutron measurements. In the present study, neutron spectrometry was investigated for verification of UF6 enrichment in 30B storage cylinders based on an unattended and passive measurement approach. MCNP5 and Geant4 simulated neutron spectra, for selected UF6 enrichments and filling profiles, were used in the investigation. The simulated neutron spectra were analyzed using principal component analysis (PCA). The PCA technique is a well-established technique and has a wide area of application including feature analysis, outlier detection, and gamma-ray spectral analysis. Results obtained demonstrate that neutron spectrometry supported by spectral feature analysis has potential for assaying UF6 enrichment in storage cylinders. The results from the present study also showed that difficulties associated with the UF6 filling profile and observed in other unattended passive neutron measurements can possibly be overcome using the approach presented.

  12. CRAD, Conduct of Operations- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January, 2005 assessment of Conduct of Operations program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  13. CRAD, Criticality Safety- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Criticality Safety program at the Y-12 - Enriched Uranium Facility.

  14. CRAD, Emergency Management- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Emergency Management program at the Y-12 Enriched Uranium Operations Oxide Conversion Facility.

  15. CRAD, DOE Oversight- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a DOE independent oversight assessment of the Y-12 Site Office's programs for oversight of its contractors at the Y-12 Enriched Uranium Operations Oxide Conversion Facility.

  16. CRAD, Occupational Safety & Health- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Industrial Safety and Industrial Health programs at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  17. CRAD, Radiological Controls- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Radiation Protection Program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  18. CRAD, Environmental Protection- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Environmental Compliance program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  19. Neutron spectrometry for ${\\rm UF}_6$ enrichment verification in storage cylinders

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Mengesha, Wondwosen; Kiff, Scott D.

    2015-01-29

    Verification of declared UF6 enrichment and mass in storage cylinders is of great interest in nuclear material nonproliferation. Nondestructive assay (NDA) techniques are commonly used for safeguards inspections to ensure accountancy of declared nuclear materials. Common NDA techniques used include gamma-ray spectrometry and both passive and active neutron measurements. In the present study, neutron spectrometry was investigated for verification of UF6 enrichment in 30B storage cylinders based on an unattended and passive measurement approach. MCNP5 and Geant4 simulated neutron spectra, for selected UF6 enrichments and filling profiles, were used in the investigation. The simulated neutron spectra were analyzed using principalmorecomponent analysis (PCA). The PCA technique is a well-established technique and has a wide area of application including feature analysis, outlier detection, and gamma-ray spectral analysis. Results obtained demonstrate that neutron spectrometry supported by spectral feature analysis has potential for assaying UF6 enrichment in storage cylinders. The results from the present study also showed that difficulties associated with the UF6 filling profile and observed in other unattended passive neutron measurements can possibly be overcome using the approach presented.less

  20. CRAD, Safety Basis- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Safety Basis at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  1. Design of an Unattended Environmental Aerosol Sampling and Analysis System for Gaseous Centrifuge Enrichment Plants

    SciTech Connect (OSTI)

    Anheier, Norman C.; Munley, John T.; Alexander, M. L.

    2011-07-19

    The resources of the IAEA continue to be challenged by the rapid, worldwide expansion of nuclear energy production. Gaseous centrifuge enrichment plants (GCEPs) represent an especially formidable dilemma to the application of safeguard measures, as the size and enrichment capacity of GCEPs continue to escalate. During the early part of the 1990's, the IAEA began to lay the foundation to strengthen and make cost-effective its future safeguard regime. Measures under Part II of 'Programme 93+2' specifically sanctioned access to nuclear fuel production facilities and environmental sampling by IAEA inspectors. Today, the Additional Protocol grants inspection and environmental sample collection authority to IAEA inspectors at GCEPs during announced and low frequency unannounced (LFUA) inspections. During inspections, IAEA inspectors collect environmental swipe samples that are then shipped offsite to an analytical laboratory for enrichment assay. This approach has proven to be an effective deterrence to GCEP misuse, but this method has never achieved the timeliness of detection goals set forth by IAEA. Furthermore it is questionable whether the IAEA will have the resources to even maintain pace with the expansive production capacity of the modern GCEP, let alone improve the timeliness in reaching current safeguards conclusions. New safeguards propositions, outside of familiar mainstream safeguard measures, may therefore be required that counteract the changing landscape of nuclear energy fuel production. A new concept is proposed that offers rapid, cost effective GCEP misuse detection, without increasing LFUA inspection access or introducing intrusive access demands on GCEP operations. Our approach is based on continuous onsite aerosol collection and laser enrichment analysis. This approach mitigates many of the constraints imposed by the LFUA protocol, reduces the demand for onsite sample collection and offsite analysis, and overcomes current limitations associated with the in-facility misuse detection devices. Onsite environmental sample collection offers the ability to collect fleeting uranium hexafluoride emissions before they are lost to the ventilation system or before they disperse throughout the facility, to become deposited onto surfaces that are contaminated with background and historical production material. Onsite aerosol sample collection, combined with enrichment analysis, provides the unique ability to quickly detect stepwise enrichment level changes within the facility, leading to a significant strengthening of facility misuse deterence. We report in this paper our study of several GCEP environmental sample release scenarios and simulation results of a newly designed aerosol collection and particle capture system that is fully integrated with the Laser Ablation, Absorbance Ratio Spectrometry (LAARS) uranium particle enrichment analysis instrument that was developed at the Pacific Northwest National Laboratory.

  2. In-Situ Measurements of Low Enrichment Uranium Holdup Process Gas Piping at K-25 - Paper for Waste Management Symposia 2010 East Tennessee Technology Park Oak Ridge, Tennessee

    SciTech Connect (OSTI)

    Rasmussen B.

    2010-01-01

    This document is the final version of a paper submitted to the Waste Management Symposia, Phoenix, 2010, abstract BJC/OR-3280. The primary document from which this paper was condensed is In-Situ Measurement of Low Enrichment Uranium Holdup in Process Gas Piping at K-25 Using NaI/HMS4 Gamma Detection Systems, BJC/OR-3355. This work explores the sufficiency and limitations of the Holdup Measurement System 4 (HJVIS4) software algorithms applied to measurements of low enriched uranium holdup in gaseous diffusion process gas piping. HMS4 has been used extensively during the decommissioning and demolition project of the K-25 building for U-235 holdup quantification. The HMS4 software is an integral part of one of the primary nondestructive assay (NDA) systems which was successfully tested and qualified for holdup deposit quantification in the process gas piping of the K-25 building. The initial qualification focused on the measurement of highly enriched UO{sub 2}F{sub 2} deposits. The purpose of this work was to determine if that qualification could be extended to include the quantification of holdup in UO{sub 2}F{sub 2} deposits of lower enrichment. Sample field data are presented to provide evidence in support of the theoretical foundation. The HMS4 algorithms were investigated in detail and found to sufficiently compensate for UO{sub 2}F{sub 2} source self-attenuation effects, over the range of expected enrichment (4-40%), in the North and East Wings of the K-25 building. The limitations of the HMS4 algorithms were explored for a described set of conditions with respect to area source measurements of low enriched UO{sub 2}F{sub 2} deposits when used in conjunction with a 1 inch by 1/2 inch sodium iodide (NaI) scintillation detector. The theoretical limitations of HMS4, based on the expected conditions in the process gas system of the K-25 building, are related back to the required data quality objectives (DQO) for the NBA measurement system established for the K-25 demolition project. The combined review of the HMS software algorithms and supporting field measurements lead to the conclusion that the majority of process gas pipe measurements are adequately corrected for source self-attenuation using HMS4. While there will be instances where the UO{sub 2}F{sub 2} holdup mass presents an infinitely thick deposit to the NaI-HMS4 system these situations are expected to be infrequent. This work confirms that the HMS4 system can quantify UO{sub 2}F{sub 2} holdup, in its current configuration (deposition, enrichment, and geometry), below the DQO levels for the K-25 building decommissioning and demolition project. For an area measurement of process gas pipe in the K-25 building, if an infinitely thick UO{sub 2}F{sub 2} deposit is identified in the range of enrichment of {approx}4-40%, the holdup quantity exceeds the corresponding DQO established for the K-25 building demolition project.

  3. The Blend Down Monitoring System Demonstration at the Paducah Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    Benton, J.; Close, D.; Johnson, W., Jr.; Kerr, P.; March-Leuba, J.; Mastal, E.; Moss, C.; Powell, D.; Sumner, J.; Uckan, T.; Vines, R.; Wright, P.D.

    1999-07-25

    Agreements between the governments of the US and the Russian Federation for the US purchase of low enriched uranium (LEU) derived from highly enriched uranium (HEU) from dismantled Russian nuclear weapons calls for the establishment of transparency measures to provide confidence that nuclear nonproliferation goals are being met. To meet these transparency goals, the agreements call for the installation of nonintrusive US instruments to monitor the down blending of HEU to LEU. The Blend Down Monitoring System (BDMS) has been jointly developed by the Los Alamos National Laboratory (LANL) and the Oak Ridge National Laboratory (ORNL) to continuously monitor {sup 235}U enrichments and mass flow rates at Russian blending facilities. Prior to its installation in Russian facilities, the BDMS was installed and operated in a UF{sub 6} flow loop in the Paducah Gaseous Diffusion Plant simulating flow and enrichment conditions expected in a typical down-blending facility. A Russian delegation to the US witnessed the equipment demonstration in June, 1998. To conduct the demonstration in the Paducah Gaseous Diffusion Plant (PGDP), the BDMS was required to meet stringent Nuclear Regulatory Commission licensing, safety and operational requirements. The Paducah demonstration was an important milestone in achieving the operational certification for the BDMS use in Russian facilities.

  4. EIS-0471: Department of Energy Loan Guarantee to Support Proposed Eagle Rock Enrichment Facility in Bonneville County, Idaho

    Broader source: Energy.gov [DOE]

    This EIS evaluates the environmental impacts of construction, operation, and decommissioning of the proposed Eagle Rock Enrichment Facility (EREF), a gas centrifuge uranium enrichment facility to be located in a rural area in western Bonneville County, Idaho. (DOE adopted this EIS issued by NRC on 04/13/2007.)

  5. Redox Exchange Induced MnO2 Nanoparticle Enrichment in Poly(3,4-ethylenedioxythiophene) Nanowires for Electrochemical Energy Storage

    SciTech Connect (OSTI)

    Liu, R.; Duay, Jonathon; Lee, Sang Bok

    2010-06-30

    MnO2 nanoparticle enriched poly(3,4-ethylenedioxythiophene) (PEDOT) nanowires are fabricated by simply soaking the PEDOT nanowires in potassium permanganate (KMnO4) solution. The structures of these MnO2 nanoparticle enriched PEDOT nanowires are characterized by SEM and TEM, which show that the MnO2 nanoparticles have uniform sizes and are finely dispersed in the PEDOT matrix. The chemical constituents and bonding of these composite nanowires are characterized by energy-dispersive X-ray analysis, X-ray photoelectron spectroscopy, and infrared spectroscopy, which indicate that the formation and dispersion of these MnO2 nanoparticles into the nanoscale pores of the PEDOT nanowires are most likely triggered by the reduction of KMnO4 via the redox exchange of permanganate ions with the functional group on PEDOT. Varying the concentrations of KMnO4 and the reaction time controls the loading amount and size of the MnO2 nanoparticles. Cyclic voltammetry and galvanostatic charge-discharge are used to characterize the electrochemical properties of these MnO2 nanoparticle loaded PEDOT nanowires. Due to their extremely high exposed surface area with nanosizes, the pristine MnO2 nanoparticles in these MnO2 nanoparticle enriched PEDOT nanowires show very high specific capacitance (410 F/g) as the supercapacitor electrode materials as well as high Li+ storage capacity (300 mAh/g) as cathode materials of Li ion battery, which boost the energy storage capacity of PEDOT nanowires to 4 times without causing excessive volume expansion in the polymer. The highly conductive and porous PEDOT matrix facilitates fast charge/discharge of the MnO2 nanoparticles and prevents them from agglomerating. These synergic properties enable the MnO2 nanoparticle enriched PEDOT nanowires to be promising electrode materials for supercapacitors and lithium ion batteries.

  6. Microclimatic performance of a free-air warming and CO₂ enrichment experiment in windy Wyoming, USA

    SciTech Connect (OSTI)

    LeCain, Daniel; Smith, David; Morgan, Jack; Kimball, Bruce A.; Pendall, Elise; Miglietta, Franco; Liang, Wenju

    2015-02-06

    In order to plan for global changing climate experiments are being conducted in many countries, but few have monitored the effects of the climate change treatments (warming, elevated CO₂) on the experimental plot microclimate. During three years of an eight year study with year-round feedback-controlled infra-red heater warming (1.5/3.0°C day/night) and growing season free-air CO₂ enrichment (600 ppm) in the mixed-grass prairie of Wyoming, USA, we monitored soil, leaf, canopy-air, above-canopy-air temperatures and relative humidity of control and treated experimental plots and evaluated ecologically important temperature differentials. Leaves were warmed somewhat less than the target settings (1.1 & 1.5°C day/night) but soil was warmed more creating an average that matched the target settings extremely well both during the day and night plus the summer and winter. The site typically has about 50% bare or litter covered soil, therefore soil heat transfer is more critical than in dense canopy ecosystems. The Wyoming site commonly has strong winds (5 ms⁻¹ average) and significant daily and seasonal temperature fluctuations (as much as 30°C daily) but the warming system was nearly always able to maintain the set temperatures regardless of abiotic variation. The within canopy-air was only slightly warmed and above canopy-air was not warmed by the system, therefore convective warming was minor. Elevated CO₂ had no direct effect nor interaction with the warming treatment on microclimate. Relative humidity within the plant canopy was only slightly reduced by warming. Soil water content was reduced by warming but increased by elevated CO₂. This study demonstrates the importance of monitoring the microclimate in manipulative field global change experiments so that critical physiological and ecological conclusions can be determined. Highly variable energy demand fluctuations showed that passive IR heater warming systems will not maintain desired warming for much of the time.

  7. Microclimatic performance of a free-air warming and CO₂ enrichment experiment in windy Wyoming, USA

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    LeCain, Daniel; Smith, David; Morgan, Jack; Kimball, Bruce A.; Pendall, Elise; Miglietta, Franco; Liang, Wenju

    2015-02-06

    In order to plan for global changing climate experiments are being conducted in many countries, but few have monitored the effects of the climate change treatments (warming, elevated CO₂) on the experimental plot microclimate. During three years of an eight year study with year-round feedback-controlled infra-red heater warming (1.5/3.0°C day/night) and growing season free-air CO₂ enrichment (600 ppm) in the mixed-grass prairie of Wyoming, USA, we monitored soil, leaf, canopy-air, above-canopy-air temperatures and relative humidity of control and treated experimental plots and evaluated ecologically important temperature differentials. Leaves were warmed somewhat less than the target settings (1.1 & 1.5°C day/night)more » but soil was warmed more creating an average that matched the target settings extremely well both during the day and night plus the summer and winter. The site typically has about 50% bare or litter covered soil, therefore soil heat transfer is more critical than in dense canopy ecosystems. The Wyoming site commonly has strong winds (5 ms⁻¹ average) and significant daily and seasonal temperature fluctuations (as much as 30°C daily) but the warming system was nearly always able to maintain the set temperatures regardless of abiotic variation. The within canopy-air was only slightly warmed and above canopy-air was not warmed by the system, therefore convective warming was minor. Elevated CO₂ had no direct effect nor interaction with the warming treatment on microclimate. Relative humidity within the plant canopy was only slightly reduced by warming. Soil water content was reduced by warming but increased by elevated CO₂. This study demonstrates the importance of monitoring the microclimate in manipulative field global change experiments so that critical physiological and ecological conclusions can be determined. Highly variable energy demand fluctuations showed that passive IR heater warming systems will not maintain desired warming for much of the time.« less

  8. The ORSphere Benchmark Evaluation and Its Potential Impact on Nuclear Criticality Safety

    SciTech Connect (OSTI)

    John D. Bess; Margaret A. Marshall; J. Blair Briggs

    2013-10-01

    In the early 1970’s, critical experiments using an unreflected metal sphere of highly enriched uranium (HEU) were performed with the focus to provide a “very accurate description…as an ideal benchmark for calculational methods and cross-section data files.” Two near-critical configurations of the Oak Ridge Sphere (ORSphere) were evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). The results from those benchmark experiments were then compared with additional unmoderated and unreflected HEU metal benchmark experiment configurations currently found in the ICSBEP Handbook. For basic geometries (spheres, cylinders, and slabs) the eigenvalues calculated using MCNP5 and ENDF/B-VII.0 were within 3 of their respective benchmark values. There appears to be generally a good agreement between calculated and benchmark values for spherical and slab geometry systems. Cylindrical geometry configurations tended to calculate low, including more complex bare HEU metal systems containing cylinders. The ORSphere experiments do not calculate within their 1s uncertainty and there is a possibility that the effect of the measured uncertainties for the GODIVA I benchmark may need reevaluated. There is significant scatter in the calculations for the highly-correlated ORCEF cylinder experiments, which are constructed from close-fitting HEU discs and annuli. Selection of a nuclear data library can have a larger impact on calculated eigenvalue results than the variation found within calculations of a given experimental series, such as the ORCEF cylinders, using a single nuclear data set.

  9. Integrated Information Technology Framework for Analysis of Data from Enrichment Plants to Support the Safeguards Mission

    SciTech Connect (OSTI)

    Marr, Clifton T.; Thurman, David A.; Jorgensen, Bruce V.

    2008-07-15

    ABSTRACT Many examples of software architectures exist that support process monitoring and analysis applications which could be applied to enrichment plants in a fashion that supports the Safeguards Mission. Pacific Northwest National Laboratory (PNNL) has developed mature solutions that will provide the framework to support online statistical analysis of enrichment plans and the entire nuclear fuel cycle. Most recently, PNNL has developed a refined architecture and supporting tools that address many of the common problems analysis and modeling environments experience: pipelining, handling large data volumes, and real-time performance. We propose the architecture and tools may be successfully used in furthering the goals of nuclear material control and accountability as both an aid to processing plant owners and as comprehensive monitoring for oversight teams.

  10. Enrichment, isolation and characterization of fungi tolerant to 1-ethyl-3-methylimidazolium acetate

    SciTech Connect (OSTI)

    Singer, S.W.; Reddy, A. P.; Gladden, J. M.; Guo, H.; Hazen, T.C.; Simmons, B. A.; VanderGheynst, J. S.

    2010-12-15

    This work aims to characterize microbial tolerance to 1-ethyl-3-methylimidazolium acetate ([C2mim][OAc]), ionic liquid that has emerged as a novel biomass pretreatment for lignocellulosic biomass. Enrichment experiments performed using inocula treated with [C2mim][OAc] under solid and liquid cultivation yielded fungal populationsdominated by Aspergilli. Ionic liquid-tolerant Aspergillus isolates from these enrichments were capable of growing in a radial plate growth assay in the presence of 10% [C2mim][OAc]. When a [C2mim][OAc]-tolerant Aspergillus fumigatus strain was grown in the presence of switchgrass, endoglucanases and xylanases were secreted that retained residual enzymatic activity in the presence of 20% [C2mim][OAc]. The results of the study suggest tolerance to ionic liquids is a general property of Aspergilli. Tolerance to an industrially important ionic liquid was discovered in a fungal genera that is widely used in biotechnology, including biomass deconstruction.

  11. A Segmented, Enriched N-type Germanium Detector for Neutrinoless Double Beta-Decay Experiments

    SciTech Connect (OSTI)

    Leviner, L.; Aalseth, Craig E.; Ahmed, M. W.; Avignone, F. T.; Back, Henning O.; Barabash, Alexander S.; Boswell, M.; De Braeckeleer, L.; Brudanin, V.; Chan, Yuen-Dat; Egorov, Viatcheslav; Elliott, Steven R.; Gehman, Victor M.; Hossbach, Todd W.; Kephart, Jeremy; Kidd, M. F.; Konovalov, S.; Lesko, Kevin; Li, Jingyi; Mei, Dongming; Mikhailov, S.; Miley, Harry S.; Radford, D. C.; Reeves, James H.; Sandukovsky, Viatcheslav; Umatov, Valdimir; Underwood, T. A.; Tornow, W.; Wu, Y. K.; Young, A.

    2014-01-21

    We present data characterizing the performance of the _rst segmented, N- type Ge detector, isotopically enriched to 85% 76Ge. This detector, based on the Ortec PT6x2 design and referred to as SEGA (Segmented, Enriched Germanium Assembly), was developed as a possible prototype for neutrinoless double beta-decay measurements by the Majorana collaboration. We present some of the general characteristics (including bias potential, efficiency, leakage current, and integral cross-talk) for this detector in its temporary cryostat. We also present an analysis of the resolution of the detector, and demonstrate that for all but two segments there is at least one channel that reaches the Majorana resolution goal below 4 keV FWHM at 2039 keV, and all channels are below 4.5 keV FWHM.

  12. EIS-0279-SA-01: Supplement Analysis | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    -SA-01: Supplement Analysis EIS-0279-SA-01: Supplement Analysis Savannah River Site Spent Nuclear Fuel Management (DOE/EIS-0279-SA-01 and DOE/EIS-0218-SA-06) This Supplement Analysis evaluates DOE's proposal to change the management method for approximately 3.3 metric tons of heavy metal (MTHM) of aluminum-clad spent nuclear fuel (SNF) from melt and dilute to conventional processing in H-Canyon at the Savannah River Site (SRS) and to down-blend the resultant highly-enriched uranium (HEU) to low

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

    SciTech Connect (OSTI)

    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.

  14. gtri | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Home gtri Joint Statement by the United States and Italy on the 2014 Nuclear Security Summit See a fact sheet here. The White HouseOffice of the Press Secretary Italy and the United States of America are pleased to announce that they have jointly completed the removal of approximately 20 kilograms of excess highly enriched uranium (HEU) and separated plutonium from Italy. At the 2012... Joint Statement by President Obama and Prime Minister Elio Di Rupo of Belgium on the 2014 Nuclear Security

  15. Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990

    SciTech Connect (OSTI)

    Wiencek, T.C.

    1995-08-01

    This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched ({approx}93% {sup 235}U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm{sup 3}) HEU fuel elements to highly loaded (up to 7 g U/cm{sup 3}) low-enrichment (<20% {sup 235}U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing.

  16. Studies of Plutonium-238 Production at the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Lastres, Oscar; Chandler, David; Jarrell, Joshua J; Maldonado, G. Ivan

    2011-01-01

    The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is a versatile 85 MW{sub th}, pressurized, light water-cooled and -moderated research reactor. The core consists of two fuel elements, an inner fuel element (IFE) and an outer fuel element (OFE), each constructed of involute fuel plates containing high-enriched-uranium (HEU) fuel ({approx}93 wt% {sup 235}U/U) in the form of U{sub 3}O{sub 8} in an Al matrix and encapsulated in Al-6061 clad. An over-moderated flux trap is located in the center of the core, a large beryllium reflector is located on the outside of the core, and two control elements (CE) are located between the fuel and the reflector. The flux trap and reflector house numerous experimental facilities which are used for isotope production, material irradiation, and cold/thermal neutron scattering. Over the past five decades, the US Department of Energy (DOE) and its agencies have been producing radioisotope power systems used by the National Aeronautics and Space Administration (NASA) for unmanned, long-term space exploration missions. Plutonium-238 is used to power Radioisotope Thermoelectric Generators (RTG) because it has a very long half-life (t{sub 1/2} {approx} 89 yr.) and it generates about 0.5 watts/gram when it decays via alpha emission. Due to the recent shortage and uncertainty of future production, the DOE has proposed a plan to the US Congress to produce {sup 238}Pu by irradiating {sup 237}Np as early as in fiscal year 2011. An annual production rate of 1.5 to 2.0 kg of {sup 238}Pu is expected to satisfy these needs and could be produced in existing national nuclear facilities like HFIR and the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Reactors at the Savannah River Site were used in the past for {sup 238}Pu production but were shut down after the last production in 1988. The nation's {sup 237}Np inventory is currently stored at INL. A plan for producing {sup 238}Pu at US research reactor facilities such as the High Flux Isotope Reactor at ORNL has been initiated by the US DOE and NASA for space exploration needs. Two Monte Carlo-based depletion codes, TRITON (ORNL) and VESTA (IRSN), were used to study the {sup 238}Pu production rates with varying target configurations in a typical HFIR fuel cycle. Preliminary studies have shown that approximately 11 grams and within 15 to 17 grams of {sup 238}Pu could be produced in the first irradiation cycle in one small and one large VXF facility, respectively, when irradiating fresh target arrays as those herein described. Important to note is that in this study we discovered that small differences in assumptions could affect the production rates of Pu-238 observed. The exact flux at a specific target location can have a significant impact upon production, so any differences in how the control elements are modeled as a function of exposure, will also cause differences in production rates. In fact, the surface plot of the large VXF target Pu-238 production shown in Figure 3 illustrates that the pins closest to the core can potentially have production rates as high as 3 times those of pins away from the core, thus implying that a cycle-to-cycle rotation of the targets may be well advised. A methodology for generating spatially-dependent, multi-group self-shielded cross sections and flux files with the KENO and CENTRM codes has been created so that standalone ORIGEN-S inputs can be quickly constructed to perform a variety of {sup 238}Pu production scenarios, i.e. combinations of the number of arrays loaded and the number of irradiation cycles. The studies herein shown with VESTA and TRITON/KENO will be used to benchmark the standalone ORIGEN.

  17. Free Air CO2 Enrichment (FACE) Data from the Duke Forest FACE Facility

    DOE Data Explorer [Office of Scientific and Technical Information (OSTI)]

    DOE has conducted trace gas enrichment experiments since the mid 1990s. The FACE Data Management System is a central repository and archive for Free-Air Carbon Dioxide Enrichment (FACE) data, as well as for the related open-top chamber (OTC) experiments. FACE Data Management System is located at DOEs Carbon Dioxide Information Analysis Center (CDIAC). While the data from the various FACE sites, each one a unique user facility, are centralized at CDIAC, each of the FACE sites presents its own view of its activities and information. For that reason, DOE Data Explorer users are advised to see both the central repository at http://public.ornl.gov/face/index.shtml and the individual home pages of each site. The Duke University FACE website actually presents information on several FACE experiments. The Forest-Atmosphere Carbon Transfer and Storage (FACTS-I) facility is located in the Blackwood Division of the Duke Forest. It consists of four free-air CO2 enrichment (FACE) plots that provide elevated atmospheric CO2 concentration and four plots that provide ambient CO2 control. The system has been in operation since June, 1994 in the prototype plot, and since August, 1996 in the three additional plots. The prototype plot and its reference were halved with a barrier inserted in the soil in 1998 to conduct, together with five additional plot pairs, CO2 X soil nutrient enrichment experiments. The rest of the plots were partitioned in early 2005 and incorporated into the CO2 X nutrient experiment. To increase statistical power, four additional ambient plots were established in January, 2005, halved, and one half of each fertilized. [copied from http://face.env.duke.edu/description.cfm] The Duke FACE home page makes information available from both completed and ongoing projects, provides a searchable database of publications and presentations, and data, images, and links to related websites.

  18. Jefferson Lab Offers Science Enrichment Program for 5th, 6th & 8th Grade

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Teachers; Registration Deadline is Sept. 11 | Jefferson Lab 1 A highlight of the Jefferson Lab Science Activities for Teachers program is Teacher Night - held at the end of the program. During the event, JSAT participants share with hundreds of teachers, from across the area, their favorite and most effective classroom science activities. These are photos from the Teacher Night held on April 1, 2015; next year's event will be April 13, 2016. Jefferson Lab Offers Science Enrichment Program

  19. Free Air CO2 Enrichment (FACE) Data from the Duke Forest FACE Facility

    DOE Data Explorer [Office of Scientific and Technical Information (OSTI)]

    DOE has conducted trace gas enrichment experiments since the mid 1990s. The FACE Data Management System is a central repository and archive for Free-Air Carbon Dioxide Enrichment (FACE) data, as well as for the related open-top chamber (OTC) experiments. FACE Data Management System is located at DOEÆs Carbon Dioxide Information Analysis Center (CDIAC). While the data from the various FACE sites, each one a unique user facility, are centralized at CDIAC, each of the FACE sites presents its own view of its activities and information. For that reason, DOE Data Explorer users are advised to see both the central repository at http://public.ornl.gov/face/index.shtml and the individual home pages of each site. The Duke University FACE website actually presents information on several FACE experiments. The Forest-Atmosphere Carbon Transfer and Storage (FACTS-I) facility is located in the Blackwood Division of the Duke Forest. It consists of four free-air CO2 enrichment (FACE) plots that provide elevated atmospheric CO2 concentration and four plots that provide ambient CO2 control. The system has been in operation since June, 1994 in the prototype plot, and since August, 1996 in the three additional plots. The prototype plot and its reference were halved with a barrier inserted in the soil in 1998 to conduct, together with five additional plot pairs, CO2 X soil nutrient enrichment experiments. The rest of the plots were partitioned in early 2005 and incorporated into the CO2 X nutrient experiment. To increase statistical power, four additional ambient plots were established in January, 2005, halved, and one half of each fertilized. [copied from http://face.env.duke.edu/description.cfm] The Duke FACE home page makes information available from both completed and ongoing projects, provides a searchable database of publications and presentations, and data, images, and links to related websites.

  20. Energy Department Selects Global Laser Enrichment for Future Operations at Paducah Site

    Broader source: Energy.gov [DOE]

    Washington, D.C. – The U.S. Department of Energy announced today that it will open negotiations with Global Laser Enrichment (GLE) for the sale of the depleted uranium hexafluoride inventory. The Department determined that GLE offered the greatest benefit to the government among those who responded to a Request for Offers (RFO) released earlier this year. Through the RFO review process, the Department also decided to enter into negotiations with AREVA for the off-specification uranium hexafluoride inventory.

  1. Economic and Non-proliferation Policy Considerations of Uranium Enrichment in Brazil and Argentina

    SciTech Connect (OSTI)

    Short, Steven M.; Phillips, Jon R.; Weimar, Mark R.; Mahy, Heidi A.

    2008-09-01

    The nuclear development programs of both Argentina and Brazil have, since the 1970s, been premised on the desire for self-sufficiency and assurance of nuclear fuel supply. While military rivalry and mutual distrust led to nuclear weapons related development programs in the 1970s and 1980s, both countries have since terminated these programs. Furthermore, the governments of both countries have pledged their commitment to exclusively non-explosive use of nuclear energy and have signed the Non Proliferation Treaty (NPT). Utilizing rights provided for under the NPT, both Argentina and Brazil have nuclear fuel production facilities, with the notable exception of enrichment plants, that provide much of the current indigenous fuel requirements for their nuclear power plants. However, both countries are actively developing enrichment capability to fill this gap. The purpose of this report is to assess the economic basis and non-proliferation policy considerations for indigenous enrichment capability within the context of their desired self-sufficiency and to evaluate possible United States Government policy options.

  2. RADIO FREQUENCY IDENTIFICATION DEVICES: EFFECTIVENESS IN IMPROVING SAFEGUARDS AT GAS-CENTRIFUGE URANIUM-ENRICHMENT PLANTS.

    SciTech Connect (OSTI)

    JOE,J.

    2007-07-08

    Recent advances in radio frequency identification devices (RFIDs) have engendered a growing interest among international safeguards experts. Potentially, RFIDs could reduce inspection work, viz. the number of inspections, number of samples, and duration of the visits, and thus improve the efficiency and effectiveness of international safeguards. This study systematically examined the applications of RFIDs for IAEA safeguards at large gas-centrifuge enrichment plants (GCEPs). These analyses are expected to help identify the requirements and desirable properties for RFIDs, to provide insights into which vulnerabilities matter most, and help formulate the required assurance tests. This work, specifically assesses the application of RFIDs for the ''Option 4'' safeguards approach, proposed by Bruce Moran, U. S. Nuclear Regulatory Commission (NRC), for large gas-centrifuge uranium-enrichment plants. The features of ''Option 4'' safeguards include placing RFIDs on all feed, product and tails (F/P/T) cylinders, along with WID readers in all FP/T stations and accountability scales. Other features of Moran's ''Option 4'' are Mailbox declarations, monitoring of load-cell-based weighing systems at the F/P/T stations and accountability scales, and continuous enrichment monitors. Relevant diversion paths were explored to evaluate how RFIDs improve the efficiency and effectiveness of safeguards. Additionally, the analysis addresses the use of RFIDs in conjunction with video monitoring and neutron detectors in a perimeter-monitoring approach to show that RFIDs can help to detect unidentified cylinders.

  3. Safety Evaluation Report for the Claiborne Enrichment Center, Homer, Louisiana (Docket No. 70-3070)

    SciTech Connect (OSTI)

    Not Available

    1994-01-01

    This report documents the US Nuclear Regulatory Commission (NRC) staff review and safety evaluation of the Louisiana Energy Services, L.P. (LES, the applicant) application for a license to possess and use byproduct, source, and special nuclear material and to enrich natural uranium to a maximum of 5 percent U-235 by the gas centrifuge process. The plant, to be known as the Claiborne Enrichment Center (CEC), would be constructed near the town of Homer in Claiborne Parish, Louisiana. At full production in a given year, the plant will receive approximately 4,700 tonnes of feed UF{sub 6} and produce 870 tonnes of low-enriched UF{sub 6}, and 3,830 tonnes of depleted UF{sub 6} tails. Facility construction, operation, and decommissioning are expected to last 5, 30, and 7 years, respectively. The objective of the review is to evaluate the potential adverse impacts of operation of the facility on worker and public health and safety under both normal operating and accident conditions. The review also considers the management organization, administrative programs, and financial qualifications provided to assure safe design and operation of the facility. The NRC staff concludes that the applicant`s descriptions, specifications, and analyses provide an adequate basis for safety review of facility operations and that construction and operation of the facility does not pose an undue risk to public health and safety.

  4. Diesel engine experiments with oxygen enrichment, water addition and lower-grade fuel

    SciTech Connect (OSTI)

    Sekar, R.R.; Marr, W.W.; Cole, R.L.; Marciniak, T.J. ); Schaus, J.E. )

    1990-01-01

    The concept of oxygen enriched air applied to reciprocating engines is getting renewed attention in the context of the progress made in the enrichment methods and the tougher emissions regulations imposed on diesel and gasoline engines. An experimental project was completed in which a direct injection diesel engine was tested with intake oxygen levels of 21% -- 35%. Since an earlier study indicated that it is necessary to use a cheaper fuel to make the concept economically attractive, a less refined fuel was included in the test series. Since a major objection to the use of oxygen enriched combustion air had been the increase in NO{sub x} emissions, a method must be found to reduce NO{sub x}. Introduction of water into the engine combustion process was included in the tests for this purpose. Fuel emulsification with water was the means used here even though other methods could also be used. The teat data indicated a large increase in engine power density, slight improvement in thermal efficiency, significant reductions in smoke and particulate emissions and NO{sub x} emissions controllable with the addition of water. 15 refs., 10 figs., 2 tabs.

  5. ORALLOY (93.2 235U) METAL CYLINDER WITH BERYLLIUM TOP REFLECTOR

    SciTech Connect (OSTI)

    John D. Bess; Leland M. Montierth; Raymond Reed; John T. Mihalczo

    2010-09-01

    A variety of critical experiments were constructed of enriched uranium metal during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, one experiment was comprised of a stack of approximately 7-inch-diameter metal discs. The bottom of the stack consisted of uranium with an approximate height of 4-1/8 inches. The top of the stack consisted of beryllium with an approximate height of 5-9/16 inches. This experiment was performed on August 20, 1963 by J. T. Mihalczo and R. G. Taylor (Ref. 1) with accompanying logbook. Both detailed and simplified model specifications are provided in this evaluation. This fast-spectra experiment was determined to represent an acceptable benchmark. The calculated eigenvalues for both the detailed and simple models are within approximately 0.5% of the benchmark values, but significantly greater than 3s from the benchmark value because the uncertainty in the benchmark is very small: 0.0002 (1s). There is significant variability between results using different neutron cross section libraries, the greatest being a ?keff of ~0.65% . Unreflected and unmoderated experiments with the same highly enriched uranium metal parts were performed at the Oak Ridge Critical Experiments Facility in the 1960s and are evaluated in HEU MET FAST 051. Thin graphite reflected (2 inches or less) experiments also using the same highly enriched uranium metal parts are evaluated in HEU MET FAST 071. Polyethylene-reflected configurations are evaluated in HEU-MET-FAST-076. Highly enriched metal annuli with beryllium cores are evaluated in HEU-MET-FAST-059.

  6. ORALLOY (93.15 235U) METAL ANNULI WITH BERYLLIUM CORE

    SciTech Connect (OSTI)

    John D. Bess; Leland M. Montierth; Raymond L. Reed; John T. Mihalczo

    2010-09-01

    A variety of critical experiments were constructed of enriched uranium metal during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, two were performed that consisted of uranium metal annuli with a solid beryllium metal core. The outer diameter of the annuli was approximately 13 or 15 inches with an inner diameter of 7 inches. The diameter of the core was approximately 7 inches. The critical height of the configurations was approximately 5 and 4 inches, respectively. The uranium annuli consisted of multiple stacked rings with diametral thicknesses of approximately 2 inches apiece and varying heights. The 15-inch experiment was performed on June 4, 1963, and the 13-inch experiment on July 12, 1963 by J. T. Mihalczo and R. G. Taylor (Ref. 1) with accompanying logbook. Both detailed and simplified model specifications are provided in this evaluation. Both of these fast-spectra experiments were determined to represent acceptable benchmarks. The calculated eigenvalues for both the detailed and simple models are within approximately 0.6% of the benchmark values, but significantly greater than 3s from the benchmark value because the uncertainty in the benchmark is very small: <0.0004 (1s). There is significant variability between results using different neutron cross section libraries, the greatest being a ?keff of ~0.67%. Unreflected and unmoderated experiments with the same highly enriched uranium metal parts were performed at the Oak Ridge Critical Experiments Facility in the 1960s and are evaluated in HEU MET FAST 051. Thin graphite reflected (2 inches or less) experiments also using the same highly enriched uranium metal parts are evaluated in HEU MET FAST 071. Polyethylene-reflected configurations are evaluated in HEU-MET-FAST-076. A stack of highly enriched metal discs with a thick beryllium reflector is evaluated in HEU-MET-FAST-069.

  7. MCNP5 CRITICALITY VALIDATION AND BIAS FOR INTERMEDIATE ENRICHED URANIUM SYSTEMS

    SciTech Connect (OSTI)

    FINFROCK SH

    2009-12-10

    The purpose of this analysis is to validate the Monte Carlo N-Particle 5 (MCNP5) code Version 1.40 (LA-UR-03-1987, 2005) and its cross-section database for k-code calculations of intermediate enriched uranium systems on INTEL{reg_sign} processor based PC's running any version of the WINDOWS operating system. Configurations with intermediate enriched uranium were modeled with the moderator range of 39 {le} H/Fissile {le} 1438. See Table 2-1 for brief descriptions of selected cases and Table 3-1 for the range of applicability for this validation. A total of 167 input cases were evaluated including bare and reflected systems in a single body or arrays. The 167 cases were taken directly from the previous (Version 4C [Lan 2005]) validation database. Section 2.0 list data used to calculate k-effective (k{sub eff}) for the 167 experimental criticality benchmark cases using the MCNP5 code v1.40 and its cross section database. Appendix B lists the MCNP cross-section database entries validated for use in evaluating the intermediate enriched uranium systems for criticality safety. The dimensions and atom densities for the intermediate enriched uranium experiments were taken from NEA/NSC/DOC(95)03, September 2005, which will be referred to as the benchmark handbook throughout the report. For these input values, the experimental benchmark k{sub eff} is approximately 1.0. The MCNP validation computer runs ran to an accuracy of approximately {+-} 0.001. For the cases where the reported benchmark k{sub eff} was not equal to 1.0000 the MCNP calculational results were normalized. The difference between the MCNP validation computer runs and the experimentally measured k{sub eff} is the MCNP5 v1.40 bias. The USLSTATS code (ORNL 1998) was utilized to perform the statistical analysis and generate an acceptable maximum k{sub eff} limit for calculations of the intermediate enriched uranium type systems.

  8. Material protection control and accounting program activities at the electrochemical plant

    SciTech Connect (OSTI)

    McAllister, S.

    1997-11-14

    The Electrochemical Plant (ECP) is the one of the Russian Federation`s four uranium enrichment plants and one of three sites in Russia blending high enriched uranium (HEU) into commercial grade low enriched uranium. ECP is located approximately 200 km east of Krasnoyarsk in the closed city of Zelenogorsk (formerly Krasnoyarsk- 45). DOE`s MPC&A program first met with ECP in September of 1996. The six national laboratories participating in DOE`s Material Protection Control and Accounting program are cooperating with ECP to enhance the capabilities of the physical protection, access control, and nuclear material control and accounting systems. The MPC&A work at ECP is expected to be completed during fiscal year 2001.

  9. Material protection control and accounting program activities at the Urals electrochemical integrated plant

    SciTech Connect (OSTI)

    McAllister, S.

    1997-11-14

    The Urals Electrochemical Integrated Plant (UEIP) is the Russian Federation`s largest uranium enrichment plant and one of three sites in Russia blending high enriched uranium (HEU) into commercial grade low enriched uranium. UEIP is located approximately 70 km north of Yekaterinburg in the closed city of Novouralsk (formerly Sverdlovsk- 44). DOE`s MPC&A program first met with UEIP in June of 1996, however because of some contractual issues the work did not start until September of 1997. The six national laboratories participating in DOE`s Material Protection Control and Accounting program are cooperating with UEIP to enhance the capabilities of the physical protection, access control, and nuclear material control and accounting systems. The MPC&A work at UEIP is expected to be completed during fiscal year 2001.

  10. Production Of High Specific Activity Copper-67

    DOE Patents [OSTI]

    Jamriska, Sr., David J.; Taylor, Wayne A.; Ott, Martin A.; Fowler, Malcolm; Heaton, Richard C.

    2003-10-28

    A process for the selective production and isolation of high specific activity Cu.sup.67 from proton-irradiated enriched Zn.sup.70 target comprises target fabrication, target irradiation with low energy (<25 MeV) protons, chemical separation of the Cu.sup.67 product from the target material and radioactive impurities of gallium, cobalt, iron, and stable aluminum via electrochemical methods or ion exchange using both anion and cation organic ion exchangers, chemical recovery of the enriched Zn.sup.70 target material, and fabrication of new targets for re-irradiation is disclosed.

  11. Production Of High Specific Activity Copper-67

    DOE Patents [OSTI]

    Jamriska, Sr., David J.; Taylor, Wayne A.; Ott, Martin A.; Fowler, Malcolm; Heaton, Richard C.

    2002-12-03

    A process for the selective production and isolation of high specific activity cu.sup.67 from proton-irradiated enriched Zn.sup.70 target comprises target fabrication, target irradiation with low energy (<25 MeV) protons, chemical separation of the Cu.sup.67 product from the target material and radioactive impurities of gallium, cobalt, iron, and stable aluminum via electrochemical methods or ion exchange using both anion and cation organic ion exchangers, chemical recovery of the enriched Zn.sup.70 target material, and fabrication of new targets for re-irradiation is disclosed.

  12. EA-1123: Transfer of Normal and Low-Enriched Uranium Billets to the United Kingdom, Hanford Site, Richland, Washington

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts of the proposal to transfer approximately 710,000 kilograms (1,562,000 pounds) of unneeded normal and low-enriched uranium to the United Kingdom; thus,...

  13. Beyond Basic Target Enrichment: New Tools to Fuel Your NGS Research ( 7th Annual SFAF Meeting, 2012)

    SciTech Connect (OSTI)

    Carter, Jennifer

    2012-06-01

    Jennifer Carter on "Beyond Basic Target Enrichment: New Tools to fuel your NGS Research" at the 2012 Sequencing, Finishing, Analysis in the Future Meeting held June 5-7, 2012 in Santa Fe, New Mexico.

  14. From the Lab to the real world : sources of error in UF {sub 6} gas enrichment monitoring

    SciTech Connect (OSTI)

    Lombardi, Marcie L.

    2012-03-01

    Safeguarding uranium enrichment facilities is a serious concern for the International Atomic Energy Agency (IAEA). Safeguards methods have changed over the years, most recently switching to an improved safeguards model that calls for new technologies to help keep up with the increasing size and complexity of today’s gas centrifuge enrichment plants (GCEPs). One of the primary goals of the IAEA is to detect the production of uranium at levels greater than those an enrichment facility may have declared. In order to accomplish this goal, new enrichment monitors need to be as accurate as possible. This dissertation will look at the Advanced Enrichment Monitor (AEM), a new enrichment monitor designed at Los Alamos National Laboratory. Specifically explored are various factors that could potentially contribute to errors in a final enrichment determination delivered by the AEM. There are many factors that can cause errors in the determination of uranium hexafluoride (UF{sub 6}) gas enrichment, especially during the period when the enrichment is being measured in an operating GCEP. To measure enrichment using the AEM, a passive 186-keV (kiloelectronvolt) measurement is used to determine the {sup 235}U content in the gas, and a transmission measurement or a gas pressure reading is used to determine the total uranium content. A transmission spectrum is generated using an x-ray tube and a “notch” filter. In this dissertation, changes that could occur in the detection efficiency and the transmission errors that could result from variations in pipe-wall thickness will be explored. Additional factors that could contribute to errors in enrichment measurement will also be examined, including changes in the gas pressure, ambient and UF{sub 6} temperature, instrumental errors, and the effects of uranium deposits on the inside of the pipe walls will be considered. The sensitivity of the enrichment calculation to these various parameters will then be evaluated. Previously, UF{sub 6} gas enrichment monitors have required empty pipe measurements to accurately determine the pipe attenuation (the pipe attenuation is typically much larger than the attenuation in the gas). This dissertation reports on a method for determining the thickness of a pipe in a GCEP when obtaining an empty pipe measurement may not be feasible. This dissertation studies each of the components that may add to the final error in the enrichment measurement, and the factors that were taken into account to mitigate these issues are also detailed and tested. The use of an x-ray generator as a transmission source and the attending stability issues are addressed. Both analytical calculations and experimental measurements have been used. For completeness, some real-world analysis results from the URENCO Capenhurst enrichment plant have been included, where the final enrichment error has remained well below 1% for approximately two months.

  15. Beyond Basic Target Enrichment: New Tools to Fuel Your NGS Research ( 7th Annual SFAF Meeting, 2012)

    ScienceCinema (OSTI)

    Carter, Jennifer [Agilent

    2013-03-22

    Jennifer Carter on "Beyond Basic Target Enrichment: New Tools to fuel your NGS Research" at the 2012 Sequencing, Finishing, Analysis in the Future Meeting held June 5-7, 2012 in Santa Fe, New Mexico.

  16. Free Air CO2 Enrichment (FACE) Research Data from the Nevada Desert FACE Facility (NDFF)

    DOE Data Explorer [Office of Scientific and Technical Information (OSTI)]

    DOE has conducted trace gas enrichment experiments since the mid 1990s. The FACE Data Management System is a central repository and archive for Free-Air Carbon Dioxide Enrichment (FACE) data, as well as for the related open-top chamber (OTC) experiments. FACE Data Management System is located at the Carbon Dioxide Information Analysis Center (CDIAC). While the data from the various FACE sites, each one a unique user facility, are centralized at CDIAC, each of the FACE sites presents its own view of its activities and information. For that reason, DOE Data Explorer users are advised to see both the central repository at http://public.ornl.gov/face/index.shtml and the individual home pages of each site. NDFF whole-ecosystem manipulation is a flagship experiment of the Terrestrial Carbon Process (TCP) research program of the US Dept. of Energy. It is also a core project of the International Geosphere-Biosphere Program (IGBP) and a contribution to the US Global Change Research Program. The NDFF was developed in conjunction with the National Science Foundation (NSF) and DOE-EPSCoR programs. FACE (Free-Air-Carbon dioxide-Enrichment) technology allows researchers to elevate the carbon dioxide level in large study plots while minimizing ecosystem disturbance. At the NDFF the concentration of CO2 was elevated by 50 percent above the present atmospheric levels in three plots in the Mojave Desert ecosystem, while six other plots remained at the current level. This experimental design provided a large area in which integrated teams of scientists could describe and quantify processes regulating carbon, nutrient, and water balances in desert ecosystems.

  17. An enriched finite element model with q-refinement for radiative boundary layers in glass cooling

    SciTech Connect (OSTI)

    Mohamed, M. Shadi; Seaid, Mohammed; Trevelyan, Jon; Laghrouche, Omar

    2014-02-01

    Radiative cooling in glass manufacturing is simulated using the partition of unity finite element method. The governing equations consist of a semi-linear transient heat equation for the temperature field and a stationary simplified P{sub 1} approximation for the radiation in non-grey semitransparent media. To integrate the coupled equations in time we consider a linearly implicit scheme in the finite element framework. A class of hyperbolic enrichment functions is proposed to resolve boundary layers near the enclosure walls. Using an industrial electromagnetic spectrum, the proposed method shows an immense reduction in the number of degrees of freedom required to achieve a certain accuracy compared to the conventional h-version finite element method. Furthermore the method shows a stable behaviour in treating the boundary layers which is shown by studying the solution close to the domain boundaries. The time integration choice is essential to implement a q-refinement procedure introduced in the current study. The enrichment is refined with respect to the steepness of the solution gradient near the domain boundary in the first few time steps and is shown to lead to a further significant reduction on top of what is already achieved with the enrichment. The performance of the proposed method is analysed for glass annealing in two enclosures where the simplified P{sub 1} approximation solution with the partition of unity method, the conventional finite element method and the finite difference method are compared to each other and to the full radiative heat transfer as well as the canonical Rosseland model.

  18. Von Hippel, at PPPL, calls for international control of nuclear enrichment

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    | Princeton Plasma Physics Lab Von Hippel, at PPPL, calls for international control of nuclear enrichment By Jeanne Jackson DeVoe February 4, 2016 Tweet Widget Google Plus One Share on Facebook Frank von Hippel discusses the Iran nuclear deal at the Jan. 30 Ronald E. Hatcher Science on Saturday lecture. (Photo by Elle Starkman/PPL Office of Communications ) Frank von Hippel discusses the Iran nuclear deal at the Jan. 30 Ronald E. Hatcher Science on Saturday lecture. Gallery: Von Hippel

  19. Jefferson Lab Offers Science Enrichment Program for 5th, 6th & 8th Grade

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Teachers | Jefferson Lab Jefferson Lab Offers Science Enrichment Program for 5th, 6th & 8th Grade Teachers Teachers Night To Improve Science Education - A highlight of the JSAT program is the annual Teacher Night, when current and former JSAT participants share with other teachers, some of the new techniques and activities they've learned or developed to enhance science education for their students. The 2013 Teacher Night will be held on April 17. NEWPORT NEWS, Va., July 9, 2012 - The

  20. Moderation control in low enriched {sup 235}U uranium hexafluoride packaging operations and transportation

    SciTech Connect (OSTI)

    Dyer, R.H.; Kovac, F.M.; Pryor, W.A.

    1993-10-01

    Moderation control is the basic parameter for ensuring nuclear criticality safety during the packaging and transport of low {sup 235}U enriched uranium hexafluoride before its conversion to nuclear power reactor fuel. Moderation control has permitted the shipment of bulk quantities in large cylinders instead of in many smaller cylinders and, therefore, has resulted in economies without compromising safety. Overall safety and uranium accountability have been enhanced through the use of the moderation control. This paper discusses moderation control and the operating procedures to ensure that moderation control is maintained during packaging operations and transportation.