Sample records for isotope reactor hfir

  1. High Flux Isotope Reactor (HFIR) | U.S. DOE Office of Science...

    Office of Science (SC) Website

    (SUF) Division SUF Home About User Facilities User Facilities Dev X-Ray Light Sources Neutron Scattering Facilities High Flux Isotope Reactor (HFIR) Lujan Neutron Scattering...

  2. High Flux Isotopes Reactor (HFIR) Cooling Towers Demolition Waste Management

    SciTech Connect (OSTI)

    Pudelek, R. E.; Gilbert, W. C.

    2002-02-26T23:59:59.000Z

    This paper describes the results of a joint initiative between Oak Ridge National Laboratory, operated by UT-Battelle, and Bechtel Jacobs Company, LLC (BJC) to characterize, package, transport, treat, and dispose of demolition waste from the High Flux Isotope Reactor (HFIR), Cooling Tower. The demolition and removal of waste from the site was the first critical step in the planned HFIR beryllium reflector replacement outage scheduled. The outage was scheduled to last a maximum of six months. Demolition and removal of the waste was critical because a new tower was to be constructed over the old concrete water basin. A detailed sampling and analysis plan was developed to characterize the hazardous and radiological constituents of the components of the Cooling Tower. Analyses were performed for Resource Conservation and Recovery Act (RCRA) heavy metals and semi-volatile constituents as defined by 40 CFR 261 and radiological parameters including gross alpha, gross beta, gross gamma, alpha-emitting isotopes and beta-emitting isotopes. Analysis of metals and semi-volatile constituents indicated no exceedances of regulatory limits. Analysis of radionuclides identified uranium and thorium and associated daughters. In addition 60Co, 99Tc, 226Rm, and 228Rm were identified. Most of the tower materials were determined to be low level radioactive waste. A small quantity was determined not to be radioactive, or could be decontaminated. The tower was dismantled October 2000 to January 2001 using a detailed step-by-step process to aid waste segregation and container loading. The volume of waste as packaged for treatment was approximately 1982 cubic meters (70,000 cubic feet). This volume was comprised of plastic ({approx}47%), wood ({approx}38%) and asbestos transite ({approx}14%). The remaining {approx}1% consisted of the fire protection piping (contaminated with lead-based paint) and incidental metal from conduit, nails and braces/supports, and sludge from the basin. The waste, except for the asbestos, was volume reduced via a private contract mechanism established by BJC. After volume reduction, the waste was packaged for rail shipment. This large waste management project successfully met cost and schedule goals.

  3. Neutron Scattering Science User Office, neutronusers@ornl.gov or (865) 574-4600. Proposals for beam time at Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR)

    E-Print Network [OSTI]

    Pennycook, Steve

    Neutron Scattering Science User Office, neutronusers@ornl.gov or (865) 574-4600. Proposals for beam Wildgruber, wildgrubercu@ornl.gov. VISION CallforProposals neutrons.ornl.gov Neutron Scattering Science - Oak time at Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR) and Spallation Neutron Source

  4. Meeting notes of the High Flux Isotope Reactor (HFIR) futures group

    SciTech Connect (OSTI)

    Houser, M.M. [comp.

    1995-08-01T23:59:59.000Z

    This report is a compilation of the notes from the ten meetings. The group charter is: (1) to identify and characterize the range of possibilities and necessities for keeping the HFIR operating for at least the next 15 years; (2) to identify and characterize the range of possibilities for enhancing the scientific and technical utility of the HFIR; (3) to evaluate the benefits or impacts of these possibilities on the various scientific fields that use the HFIR or its products; (4) to evaluate the benefits or impacts on the operation and maintenance of the HFIR facility and the regulatory requirements; (5) to estimate the costs, including operating costs, and the schedules, including downtime, for these various possibilities; and one possible impact of proposed changes may be to stimulate increased pressure for a reduced enrichment fuel for HFIR.

  5. COMSOL-based Nuclear Reactor Kinetics Studies at the HFIR

    SciTech Connect (OSTI)

    Chandler, David [ORNL] [ORNL; Freels, James D [ORNL] [ORNL; Maldonado, G Ivan [ORNL] [ORNL; Primm, Trent [ORNL] [ORNL

    2011-01-01T23:59:59.000Z

    The computational ability to accurately predict the dynamic behavior of a nuclear reactor core in response to reactivity-induced perturbations is an important subject in reactor physics. Space-time and point kinetics methodologies were developed for the purpose of studying the transient-induced behavior of the High Flux Isotope Reactor s (HFIR) compact core. The space-time simulations employed the three-energy-group neutron diffusion equations, and transients initiated by control cylinder and hydraulic tube rabbit ejections were studied. The work presented here is the first step towards creating a comprehensive multiphysics methodology for studying the dynamic behavior of the HFIR core during reactivity perturbations. The results of these studies show that point kinetics is adequate for small perturbations in which the power distribution is assumed to be time-independent, but space-time methods must be utilized to determine localized effects.

  6. HFIR spent fuel management alternatives

    SciTech Connect (OSTI)

    Begovich, J.M.; Green, V.M.; Shappert, L.B.; Lotts, A.L.

    1992-10-15T23:59:59.000Z

    The High Flux Isotope Reactor (HFIR) at Martin Marietta Energy Systems' Oak Ridge National Laboratory (ORNL) has been unable to ship its spent fuel to Savannah River Site (SRS) for reprocessing since 1985. The HFIR storage pools are expected to fill up in the February 1994 to February 1995 time frame. If a management altemative to existing HFIR pool storage is not identified and implemented before the HFIR pools are full, the HFIR will be forced to shut down. This study investigated several alternatives for managing the HFIR spent fuel, attempting to identify options that could be implemented before the HFIR pools are full. The options investigated were: installing a dedicated dry cask storage facility at ORNL, increasing HFIR pool storage capacity by clearing the HFIR pools of debris and either close-packing or stacking the spent fuel elements, storing the spent fuel at another ORNL pool, storing the spent fuel in one or more hot cells at ORNL, and shipping the spent fuel offsite for reprocessing or storage elsewhere.

  7. HFIR spent fuel management alternatives

    SciTech Connect (OSTI)

    Begovich, J.M.; Green, V.M.; Shappert, L.B.; Lotts, A.L.

    1992-10-15T23:59:59.000Z

    The High Flux Isotope Reactor (HFIR) at Martin Marietta Energy Systems` Oak Ridge National Laboratory (ORNL) has been unable to ship its spent fuel to Savannah River Site (SRS) for reprocessing since 1985. The HFIR storage pools are expected to fill up in the February 1994 to February 1995 time frame. If a management altemative to existing HFIR pool storage is not identified and implemented before the HFIR pools are full, the HFIR will be forced to shut down. This study investigated several alternatives for managing the HFIR spent fuel, attempting to identify options that could be implemented before the HFIR pools are full. The options investigated were: installing a dedicated dry cask storage facility at ORNL, increasing HFIR pool storage capacity by clearing the HFIR pools of debris and either close-packing or stacking the spent fuel elements, storing the spent fuel at another ORNL pool, storing the spent fuel in one or more hot cells at ORNL, and shipping the spent fuel offsite for reprocessing or storage elsewhere.

  8. HFIR | High Flux Isotope Reactor | ORNL

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

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  9. Calculation of Heating Values for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Peterson, Joshua L [ORNL] [ORNL; Ilas, Germina [ORNL] [ORNL

    2012-01-01T23:59:59.000Z

    Calculating the amount of energy released by a fission reaction (fission Q value) and the heating rate distribution in a nuclear reactor is an important part of the safety analysis. However, these calculations can become very complex. One of the codes that can be used for this type of analyses is the Monte Carlo transport code MCNP5. Currently it is impossible to calculate the Q value and heating rate disposition for delayed beta and delayed gamma particles directly from MCNP5. The purpose of this paper is to outline a rigorous method for indirectly calculating the Q values and heating rates in the High Flux Isotope Reactor (HFIR), based on previous similar studies carried out for very high-temperature reactor configurations. This method has been applied in this study to calculate heating rates for the beginning of cycle (BOC) and end-of-cycle (EOC) states of HFIR. In addition, the BOC results obtained for HFIR are compared with corresponding results for the Advanced Test Reactor. The fission Q value for HFIR was calculated as 200.2 MeV for the BOC and 201.3 MeV for the EOC. It was also determined that 95.1% and 95.4% of the heat was deposited within the HFIR fuel plates for the BOC and EOC models, respectively. This methodology can also be used for heating rate calculations for HFIR experiments.

  10. Nuclear Transmutations in HFIR's Beryllium Reflector and Their Impact on Reactor Operation and Reflector Disposal

    SciTech Connect (OSTI)

    Chandler, David [ORNL; Maldonado, G Ivan [ORNL; Primm, Trent [ORNL; Proctor, Larry Duane [ORNL

    2012-01-01T23:59:59.000Z

    The High Flux Isotope Reactor located at the Oak Ridge National Laboratory utilizes a large cylindrical beryllium reflector that is subdivided into three concentric regions and encompasses the compact reactor core. Nuclear transmutations caused by neutron activation occur in the beryllium reflector regions, which leads to unwanted neutron absorbing and radiation emitting isotopes. During the past year, two topics related to the HFIR beryllium reflector were reviewed. The first topic included studying the neutron poison (helium-3 and lithium-6) buildup in the reflector regions and its affect on beginning-of-cycle reactivity. A new methodology was developed to predict the reactivity impact and estimated symmetrical critical control element positions as a function of outage time between cycles due to helium-3 buildup and was shown to be in better agreement with actual symmetrical critical control element position data than the current methodology. The second topic included studying the composition of the beryllium reflector regions at discharge as well as during decay to assess the viability of transporting, storing, and ultimately disposing the reflector regions currently stored in the spent fuel pool. The post-irradiation curie inventories were used to determine whether the reflector regions are discharged as transuranic waste or become transuranic waste during the decay period for disposal purposes and to determine the nuclear hazard category, which may affect the controls invoked for transportation and temporary storage. Two of the reflector regions were determined to be transuranic waste at discharge and the other region was determined to become transuranic waste in less than 2 years after being discharged due to the initial uranium content (0.0044 weight percent uranium). It was also concluded that all three of the reflector regions could be classified as nuclear hazard category 3 (potential for localized consequences only).

  11. Upgraded HFIR Fuel Element Welding System

    SciTech Connect (OSTI)

    Sease, John D [ORNL

    2010-02-01T23:59:59.000Z

    The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. In recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.

  12. Fabrication of control rods for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Sease, J.D.

    1998-03-01T23:59:59.000Z

    The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A.

  13. High Flux Isotope Reactor (HFIR) | Nuclear Science | ORNL

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

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  14. Utilization of the High Flux Isotope Reactor at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Selby, Douglas L [ORNL; Bilheux, Hassina Z [ORNL; Meilleur, Flora [ORNL; Jones, Amy [ORNL; Bailey, William Barton [ORNL; Vandergriff, David H [ORNL

    2015-01-01T23:59:59.000Z

    This paper addresses several aspects of the scientific utilization of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR). Topics to be covered will include: 1) HFIR neutron scattering instruments and the formal instrument user program; 2) Recent upgrades to the neutron scattering instrument stations at the reactor, and 3) eMod a new tool for addressing instrument modifications and providing configuration control and design process for scientific instruments at HFIR and the Spallation Neutron Source (SNS). There are 15 operating neutron instrument stations at HFIR with 12 of them organized into a formal user program. Since the last presentation on HFIR instruments at IGORR we have installed a Single Crystal Quasi-Laue Diffractometer instrument called IMAGINE; and we have made significant upgrades to HFIR neutron scattering instruments including the Cold Triple Axis Instrument, the Wide Angle Neutron Diffractometer, the Powder Diffractometer, and the Neutron Imaging station. In addition, we have initiated upgrades to the Thermal Triple Axis Instrument and the Bio-SANS cold neutron instrument detector system. All of these upgrades are tied to a continuous effort to maintain a high level neutron scattering user program at the HFIR. For the purpose of tracking modifications such as those mentioned and configuration control we have been developing an electronic system for entering instrument modification requests that follows a modification or instrument project through concept development, design, fabrication, installation, and commissioning. This system, which we call eMod, electronically leads the task leader through a series of questions and checklists that then identifies such things as ES&H and radiological issues and then automatically designates specific individuals for the activity review process. The system has been in use for less than a year and we are still working out some of the inefficiencies, but we believe that this will become a very effective tool for achieving the configuration and process control believed to be necessary for scientific instrument systems.

  15. Reactor production of sup 252 Cf and transcurium isotopes

    SciTech Connect (OSTI)

    Alexander, C.W.; Halperin, J.; Walker, R.L.; Bigelow, J.E.

    1990-01-01T23:59:59.000Z

    Berkelium, californium, einsteinium, and fermium are currently produced in the High Flux Isotope Reactor (HFIR) and recovered in the Radiochemical Engineering Development Center (REDC) at the Oak Ridge National Laboratory (ORNL). All the isotopes are used for research. In addition, {sup 252}Cf, {sup 253}Es, and {sup 255}Fm have been considered or are used for industrial or medical applications. ORNL is the sole producer of these transcurium isotopes in the western world. A wide range of actinide samples were irradiated in special test assemblies at the Fast Flux Test Facility (FFTF) at Hanford, Washington. The purpose of the experiments was to evaluate the usefulness of the two-group flux model for transmutations in the special assemblies with an eventual goal of determining the feasibility of producing macro amounts of transcurium isotopes in the FFTF. Preliminary results from the production of {sup 254g}Es from {sup 252}Cf will be discussed. 14 refs., 5 tabs.

  16. Determination of the theoretical feasibility for the transmutation of europium isotopes from high flux isotope reactor control cylinders

    SciTech Connect (OSTI)

    Elam, K.R.; Reich, W.J.

    1995-09-01T23:59:59.000Z

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) is a 100 MWth light-water research reactor designed and built in the 1960s primarily for the production of transuranic isotopes. The HFIR is equipped with two concentric cylindrical blade assemblies, known as control cylinders, that are used to control reactor power. These control cylinders, which become highly radioactive from neutron exposure, are periodically replaced as part of the normal operation of the reactor. The highly radioactive region of the control cylinders is composed of europium oxide in an aluminum matrix. The spent HFIR control cylinders have historically been emplaced in the ORNL Waste Area Grouping (WAG) 6. The control cylinders pose a potential radiological hazard due to the long lived radiotoxic europium isotopes {sup 152}Eu, {sup 154}Eu, and {sup 155}Eu. In a 1991 health evaluation of WAG 6 (ERD 1991) it was shown that these cylinders were a major component of the total radioactivity in WAG 6 and posed a potential exposure hazard to the public in some of the postulated assessment scenarios. These health evaluations, though preliminary and conservative in nature, illustrate the incentive to investigate methods for permanent destruction of the europium radionuclides. When the cost of removing the control cylinders from WAG 6, performing chemical separations and irradiating the material in HFIR are factored in, the option of leaving the control cylinders in place for decay must be considered. Other options, such as construction of an engineered barrier around the disposal silos to reduce the chance of migration, should also be analyzed.

  17. High flux isotope reactor cold source preconceptual design study report

    SciTech Connect (OSTI)

    Selby, D.L.; Bucholz, J.A.; Burnette, S.E. [and others

    1995-12-01T23:59:59.000Z

    In February 1995, the deputy director of Oak Ridge National Laboratory (ORNL) formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced Neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. The anticipated cold source will consist of a cryogenic LH{sub 2} moderator plug, a cryogenic pump system, a refrigerator that uses helium gas as a refrigerant, a heat exchanger to interface the refrigerant with the hydrogen loop, liquid hydrogen transfer lines, a gas handling system that includes vacuum lines, and an instrumentation and control system to provide constant system status monitoring and to maintain system stability. The scope of this project includes the development, design, safety analysis, procurement/fabrication, testing, and installation of all of the components necessary to produce a working cold source within an existing HFIR beam tube. This project will also include those activities necessary to transport the cold neutron beam to the front face of the present HFIR beam room. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and research and development (R and D), (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the preconceptual phase and establishes the concept feasibility. The information presented includes the project scope, the preliminary design requirements, the preliminary cost and schedule, the preliminary performance data, and an outline of the various plans for completing the project.

  18. High Flux Isotope Reactor cold neutron source reference design concept

    SciTech Connect (OSTI)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

    1998-05-01T23:59:59.000Z

    In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

  19. Awareness, Preference, Utilization, and Messaging Research for the Spallation Neutron Source and High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Bryant, Rebecca [Bryant Research, LLC; Kszos, Lynn A [ORNL

    2011-03-01T23:59:59.000Z

    Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one-on-one interviews and traditional and online focus groups with scientists. The latter include SNS, HFIR, and APS users as well as scientists at ORNL, some of whom had not yet used HFIR and/or SNS. These approaches informed development of the second phase, a quantitative online survey. The survey consisted of 16 questions and 7 demographic categorizations, 9 open-ended queries, and 153 pre-coded variables and took an average time of 18 minutes to complete. The survey was sent to 589 SNS/HFIR users, 1,819 NSLS users, and 2,587 APS users. A total of 899 individuals provided responses for this study: 240 from NSLS; 136 from SNS/HFIR; and 523 from APS. The overall response rate was 18%.

  20. 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 [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

    2011-02-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2012-03-01T23:59:59.000Z

    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.

  2. Reactivity Accountability Attributed to Reflector Poisons in the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Chandler, David [ORNL; Maldonado, G Ivan [ORNL; Primm, Trent [ORNL

    2009-12-01T23:59:59.000Z

    The objective of this study is to develop a methodology to predict the reactivity impact as a function of outage time between cycles of 3He, 6Li, and other poisons in the High Flux Isotope Reactor s (HFIR) beryllium reflector. The reactivity worth at startup of the HFIR has been incorrectly predicted in the past after the reactor has been shut-down for long periods of time. The incorrect prediction was postulated to be due to the erroneous calculation of 3He buildup in the beryllium reflector. It is necessary to develop a better estimate of the start-of-cycle symmetric critical control element positions since if the estimated and actual symmetrical critical control element positions differ by more than $1.55 in reactivity (approximately one-half inch in control element startup position), HFIR is to be shutdown and a technical evaluation is performed to resolve the discrepancy prior to restart. 3He is generated and depleted during operation, but during an outage, the depletion of 3He ceases because it is a stable isotope. 3He is born from the radioactive decay of tritium, and thus the concentration of 3He increases during shutdown. SCALE, specifically the TRITON and CSAS5 control modules including the KENO V.A, COUPLE, and ORIGEN functional modules were utilized in this study. An equation relating the down time (td) to the change in symmetric control element position was generated and validated against measurements for approximately 40 HFIR operating cycles. The newly-derived correlation was shown to improve accuracy of predictions for long periods of down time.

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

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01T23:59:59.000Z

    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.

  4. 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. [ORNL] [ORNL; Ellis, R. J. [ORNL] [ORNL; Gehin, J. C. [ORNL] [ORNL; Clarno, K. T. [ORNL] [ORNL; Williams, K. A. [ORNL] [ORNL; Moses, D. L. [ORNL] [ORNL

    2006-11-01T23:59:59.000Z

    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.

  5. Validation of a Monte Carlo based depletion methodology via High Flux Isotope Reactor HEU post-irradiation examination measurements

    SciTech Connect (OSTI)

    Chandler, David [ORNL; Maldonado, G Ivan [ORNL; Primm, Trent [ORNL

    2010-01-01T23:59:59.000Z

    The purpose of this study is to validate a Monte Carlo based depletion methodology by comparing calculated post-irradiation uranium isotopic compositions in the fuel elements of the High Flux Isotope Reactor (HFIR) core to values measured using uranium mass-spectrographic analysis. Three fuel plates were analyzed: two from the outer fuel element (OFE) and one from the inner fuel element (IFE). Fuel plates O-111-8, O-350-1, and I-417-24 from outer fuel elements 5-O and 21-O and inner fuel element 49-I, respectively, were selected for examination. Fuel elements 5-O, 21-O, and 49-1 were loaded into HFIR during cycles 4, 16, and 35, respectively (mid to late 1960s). Approximately one year after each of these elements were irradiated, they were transferred to the High Radiation Level Examination Laboratory (HRLEL) where samples from these fuel plates were sectioned and examined via uranium mass-spectrographic analysis. The isotopic composition of each of the samples was used to determine the atomic percent of the uranium isotopes. A Monte Carlo based depletion computer program, ALEPH, which couples the MCNP and ORIGEN codes, was utilized to calculate the nuclide inventory at the end-of-cycle (EOC). A current ALEPH/MCNP input for HFIR fuel cycle 400 was modified to replicate cycles 4, 16, and 35. The control element withdrawal curves and flux trap loadings were revised, as well as the radial zone boundaries and nuclide concentrations in the MCNP model. The calculated EOC uranium isotopic compositions for the analyzed plates were found to be in good agreement with measurements, which reveals that ALEPH/MCNP can accurately calculate burn-up dependent uranium isotopic concentrations for the HFIR core. The spatial power distribution in HFIR changes significantly as irradiation time increases due to control element movement. Accurate calculation of the end-of-life uranium isotopic inventory is a good indicator that the power distribution variation as a function of space and time is accurately calculated, i.e. an integral check. Hence, the time dependent heat generation source terms needed for reactor core thermal hydraulic analysis, if derived from this methodology, have been shown to be accurate for highly enriched uranium (HEU) fuel.

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

    SciTech Connect (OSTI)

    Primm, Trent [ORNL; Guida, Tracey [University of Pittsburgh

    2010-02-01T23:59:59.000Z

    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.

  7. Source Terms for HFIR Beam Tube Shielding Analyses, and a Complete Shielding Analysis of the HB-3 Tube

    SciTech Connect (OSTI)

    Bucholz, J.A.

    2000-07-01T23:59:59.000Z

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source.

  8. STARTUP REACTIVITY ACCOUNTABILITY ATTRIBUTED TO ISOTOPIC TRANSMUTATIONS IN THE IRRADIATED BERYLLIUM REFLECTOR OF THE HIGH FLUX ISTOTOPE REACTOR

    SciTech Connect (OSTI)

    Chandler, David [ORNL] [ORNL; Maldonado, G Ivan [ORNL] [ORNL; Primm, Trent [ORNL] [ORNL

    2010-01-01T23:59:59.000Z

    The objective of this study is to develop a methodology to predict the reactivity impact as a function of outage time between cycles of 3He, 6Li, and other poisons in the High Flux Isotope Reactor s (HFIR) beryllium reflector. The reactivity worth at startup of the HFIR has been incorrectly predicted in the past after the reactor has been shut-down for long periods of time. The incorrect prediction was postulated to be due to the erroneous calculation of 3He buildup in the beryllium reflector. It is necessary to develop a better estimate of the start-of-cycle symmetric critical control element positions since if the estimated and actual symmetrical critical control element positions differ by more than $1.55 in reactivity (approximately one-half inch in control element startup position), HFIR is to be shutdown and a technical evaluation is performed to resolve the discrepancy prior to restart. 3He is generated and depleted during operation, but during an outage, the depletion of 3He ceases because it is a stable isotope. 3He is born from the radioactive decay of tritium, and thus the concentration of 3He increases during shutdown. The computer program SCALE, specifically the TRITON and CSAS5 control modules including the KENO V.A, COUPLE, and ORIGEN functional modules were utilized in this study. An equation relating the down time (td) to the change in symmetric control element position was generated and validated against measurements for approximately 40 HFIR operating cycles. The newly-derived correlation was shown to improve accuracy of predictions for long periods of down time.

  9. Density of Gadolinium Nitrate Solutions for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Taylor, Paul Allen [ORNL; Lee, Denise L [ORNL

    2009-05-01T23:59:59.000Z

    In late 1992, the High Flux Isotope Reactor (HFIR) was planning to switch the solution contained in the poison injection tank from cadmium nitrate to gadolinium nitrate. The poison injection system is an emergency system used to shut down the reactor by adding a neutron poison to the cooling water. This system must be able to supply a minimum of 69 pounds of gadolinium to the reactor coolant system in order to guarantee that the reactor would become subcritical. A graph of the density of gadolinium nitrate solutions over a concentration range of 5 to 30 wt% and a temperature range of 15 to 40{sup o}C was prepared. Routine density measurements of the solution in the poison injection tank are made by HFIR personnel, and an adaptation of the original graph is used to determine the gadolinium nitrate concentration. In late 2008, HFIR personnel decided that the heat tracing that was present on the piping for the poison injection system could be removed without any danger of freezing the solution; however, the gadolinium nitrate solution might get as cold as 5{sup o}C. This was outside the range of the current density-concentration correlation, so the range needed to be expanded. This report supplies a new density-concentration correlation that covers the extended temperature range. The correlation is given in new units, which greatly simplifies the calculation that is required to determine the pounds of gadolinium in the tank solution. The procedure for calculating the amount of gadolinium in the HFIR poison injection system is as follows: (1) Calculate the usable volume in the system; (2) Measure the density of the solution; (3) Calculate the gadolinium concentration using the following equation: Gd(lb/ft{sup 3}) = measured density (g/mL) x 34.681 - 34.785; (4) Calculate the amount of gadolinium in the system using the following equation: Amount of Gd(lb) = Gd concentration (lb/ft{sup 3}) x usable volume (ft{sup 3}). The equation in step 3 is exact for a temperature of 5{sup o}C, and overestimates the gadolinium concentration at all higher temperatures. This guarantees that the calculation is conservative, in that the actual concentration will be at least as high as that calculated. If an additional safety factor is desired, it is recommended that an administrative control limit be set that is higher than the required minimum amount of gadolinium.

  10. Extraction of gadolinium from high flux isotope reactor control plates. [Alternative method

    SciTech Connect (OSTI)

    Kohring, M.W.

    1987-04-01T23:59:59.000Z

    Gadolinium-153 is an important radioisotope used in the diagnosis of various bone disorders. Recent medical and technical developments in the detection and cure of osteoporosis, a bone disease affecting an estimated 50 million people, have greatly increased the demand for this isotope. The Oak Ridge National Laboratory (ORNL) has produced /sup 153/Gd since 1980 primarily through the irradiation of a natural europium-oxide powder followed by the chemical separation of the gadolinium fraction from the europium material. Due to the higher demand for /sup 153/Gd, an alternative production method to supplement this process has been investigated. This process involves the extraction of gadolinium from the europium-bearing region of highly radioactive, spent control plates used at the High Flux Isotope Reactor (HFIR) with a subsequent re-irradiation of the extracted material for the production of the /sup 153/Gd. Based on the results of experimental and calculational analyses, up to 25 grams of valuable gadolinium (greater than or equal to60% enriched in /sup 152/Gd) resides in the europium-bearing region of the HFIR control components of which 70% is recoverable. At a specific activity yield of 40 curies of /sup 153/Gd for each gram of gadolinium re-irradiated, 700 one-curie sources can be produced from each control plate assayed.

  11. STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS

    SciTech Connect (OSTI)

    Katoh, Yutai [ORNL; Koyanagi, Takaaki [ORNL; Kiggans, Jim [ORNL; Cetiner, Nesrin [ORNL; McDuffee, Joel [ORNL

    2014-09-01T23:59:59.000Z

    Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

  12. Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Primm, Trent [ORNL; Gehin, Jess C [ORNL

    2009-04-01T23:59:59.000Z

    A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

  13. Analysis of HFIR Dosimetry Experiments Performed in Cycles 400 and 401

    SciTech Connect (OSTI)

    Remec, Igor [ORNL; Baldwin, Charles A [ORNL

    2008-09-01T23:59:59.000Z

    The High Flux Isotope Reactor (HFIR) has been in operation at Oak Ridge National Laboratory since 1966. To upgrade and enhance capabilities for neutron science research at the reactor, a larger HB-2 beam tube was installed in April of 2002. To assess, experimentally, the impact of this larger beam tube on radiation damage rates [i.e., displacement-per-atom (dpa) rates] used in vessel life extension studies, dosimetry experiments were performed from April to August 2004 during fuel cycles 400 and 401. This report documents the analysis of the dosimetry experiments and the determination of best-estimate dpa rates. These dpa rates are obtained by performing a least-squares adjustment of calculated neutron and gamma-ray fluxes and the measured responses of radiometric monitors and beryllium helium accumulation fluence monitors. The best-estimate dpa rates provided here will be used to update HFIR pressure vessel life extension studies, which determine the pressure/temperature limits for reactor operation and the HFIR pressure vessel's remaining life. All irradiation parameters given in this report correspond to a reactor power of 85 MW.

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

    SciTech Connect (OSTI)

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

    2007-11-01T23:59:59.000Z

    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.

  15. High Flux Isotope Reactor (HFIR) | U.S. DOE Office of Science (SC)

    Office of Science (SC) Website

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What'sis Taking Over OurTheBrookhaven National LaboratoryJeffrey L80's » GeorgeNeutron Scattering

  16. High Flux Isotope Reactor named Nuclear Historic Landmark | ornl...

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

    High Flux Isotope Reactor named Nuclear Historic Landmark The High Flux Isotope Reactor vessel at Oak Ridge National Laboratory resides in a pool of water illuminated by the blue...

  17. Neutronic Analysis of an Advanced Fuel Design Concept for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Xoubi, Ned [ORNL; Primm, Trent [ORNL; Maldonado, G. Ivan [University of Tennessee, Knoxville (UTK)

    2009-01-01T23:59:59.000Z

    This study presents the neutronic analysis of an advanced fuel design concept for the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) that could significantly extend the current fuel cycle length under the existing design and safety criteria. A key advantage of the fuel design herein proposed is that it would not require structural changes to the present HFIR core, in other words, maintaining the same rated power and fuel geometry (i.e., fuel plate thickness and coolant channel dimensions). Of particular practical importance, as well, is the fact that the proposed change could be justified within the bounds of the existing nuclear safety basis. The simulations herein reported employed transport theory-based and exposure-dependent eigenvalue characterization to help improve the prediction of key fuel cycle parameters. These parameters were estimated by coupling a benchmarked three-dimensional MCNP5 model of the HFIR core to the depletion code ORIGEN via the MONTEBURNS interface. The design of an advanced HFIR core with an improved fuel loading is an idea that evolved from early studies by R. D. Cheverton, formerly of ORNL. This study contrasts a modified and increased core loading of 12 kg of 235U against the current core loading of 9.4 kg. The simulations performed predict a cycle length of 39 days for the proposed fuel design, which represents a 50% increase in the cycle length in response to a 25% increase in fissile loading, with an average fuel burnup increase of {approx}23%. The results suggest that the excess reactivity can be controlled with the present design and arrangement of control elements throughout the core's life. Also, the new power distribution is comparable or even improved relative to the current power distribution, displaying lower peak to average fission rate densities across the inner fuel element's centerline and bottom cells. In fact, the fission rate density in the outer fuel element also decreased at these key locations for the proposed design. Overall, it is estimated that the advanced core design could increase the availability of the HFIR facility by {approx}50% and generate {approx}33% more neutrons annually, which is expected to yield sizeable savings during the remaining life of HFIR, currently expected to operate through 2014. This study emphasizes the neutronics evaluation of a new fuel design. Although a number of other performance parameters of the proposed design check favorably against the current design, and most of the core design features remain identical to the reference, it is acknowledged that additional evaluations would be required to fully justify the thermal-hydraulic and thermal-mechanical performance of a new fuel design, including checks for cladding corrosion performance as well as for industrial and economic feasibility.

  18. Impact of HFIR LEU Conversion on Beryllium Reflector Degradation Factors

    SciTech Connect (OSTI)

    Ilas, Dan [ORNL

    2013-10-01T23:59:59.000Z

    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. Packed bed reactor for photochemical sup 196 Hg isotope separation

    SciTech Connect (OSTI)

    Grossman, M.W.; Speer, R.

    1992-03-03T23:59:59.000Z

    This patent describes a photochemical reactor useful for the isotopic enrichment of a predetermined isotope of mercury comprising a reactor cell and a monoisotopic light source It comprises: a plurality of transparent, straight reactor cell tubes disposed axially within the internal volume of the reactor cell to increase the surface area thereof for production deposition.

  20. Lessons Learned in the Update of a Safety Limit for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Cook, David Howard [ORNL

    2009-01-01T23:59:59.000Z

    A recent unreviewed safety question (USQ) regarding a portion of the High Flux Isotope Reactor (HFIR) transient decay heat removal analysis focused on applicability of a heat transfer correlation at the low flow end of reactor operations. During resolution of this issue, review of the correlations used to establish the safety limit (SL) on reactor flux-to-flow ratio revealed the need to change the magnitude of the SL at the low flow end of reactor operations and the need to update the hot spot fuel damage criteria to incorporate current knowledge involving parallel channel flow stability. Because of the original safety design strategy for the reactor, resolution of the issues for the flux-to-flow ratio involved reevaluation of all key process variable SLs and limiting control settings (LCSs) using the current version of the heat transfer analysis code for the reactor. Goals of the work involved updating and upgrading the SL analysis where necessary, while preserving the safety design strategy for the reactor. Changes made include revisions to the safety design criteria at low flows to address the USQ, update of the process- and analysis input-variable uncertainty considerations, and upgrade of the safety design criteria at high flow. The challenges faced during update/upgrade of this SL and LCS are typical of the problems found in the integration of safety into the design process for a complex facility. In particular, the problems addressed in the area of instrument uncertainties provide valuable lessons learned for establishment and configuration control of SLs for large facilities.

  1. HFIR Experiment Facilities | ORNL Neutron Sciences

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

    Scattering Neutron Scattering Facilities at HFIR The fully instrumented HFIR will eventually include 15 state-of-the-art neutron scattering instruments, seven of which will be...

  2. HFIR History - ORNL Neutron Sciences

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

    has grown to include materials irradiation, neutron activation, and, most recently, neutron scattering. In 2007, HFIR completed the most dramatic transformation in its...

  3. 2D Thermal Hydraulic Analysis and Benchmark in Support of HFIR LEU Conversion using COMSOL

    SciTech Connect (OSTI)

    Freels, James D [ORNL; Bodey, Isaac T [ORNL; Lowe, Kirk T [ORNL; Arimilli, Rao V [ORNL

    2010-09-01T23:59:59.000Z

    The research documented herein was funded by a research contract between the Research Reactors Division (RRD) of Oak Ridge National Laboratory (ORNL) and the University of Tennessee, Knoxville (UTK) Mechanical, Aerospace and Biomedical Engineering Department (MABE). The research was governed by a statement of work (SOW) which clearly defines nine specific tasks. This report is outlined to follow and document the results of each of these nine specific tasks. The primary goal of this phase of the research is to demonstrate, through verification and validation methods, that COMSOL is a viable simulation tool for thermal-hydraulic modeling of the High Flux Isotope Reactor (HFIR) core. A secondary goal of this two-dimensional phase of the research is to establish methodology and data base libraries that are also needed in the full three-dimensional COMSOL simulation to follow. COMSOL version 3.5a was used for all of the models presented throughout this report.

  4. Irradiation effect on deuterium behaviour in low-dose HFIR neutron-irradiated tungsten

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

    Shimada, Masashi; Cao, G.; Otsuka, T.; Hara, M.; Kobayashi, M.; Oya, Y.; Hatano, Y.

    2015-01-01T23:59:59.000Z

    Tungsten samples were irradiated by neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory at reactor coolant temperatures of 50–70 ?C to low displacement damage of 0.025 and 0.3 dpa. After cooling down, the HFIR neutronirradiated tungsten samples were exposed to deuterium plasmas in the Tritium Plasma Experiment, Idaho National Laboratory at 100, 200 and 500 ?C twice at the ion fluence of 5×1025 m-2 to reach the total ion fluence of 1×1026 m-2 in order to investigate maximum near-surface (more »in 0.3 dpa samples. The large discrepancy between the total retention via thermal desorption spectroscopy and the nearsurface retention via nuclear reaction analysis indicated the deuterium was trapped in bulk (at least 50µm depth for 0.025 dpa and 35µm depth for 0.3 dpa) at 500 ?C cases even in the relatively low ion fluence of 1026 m-2.« less

  5. Irradiation effect on deuterium behaviour in low-dose HFIR neutron-irradiated tungsten

    SciTech Connect (OSTI)

    Shimada, Masashi [Idaho National Lab. (INL), Idaho Falls, ID (United States).Fusion Safety Program; Cao, G. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Otsuka, T. [Kyushu Univ., Fukuoka (Japan). Interdisciplinary Graduate School of Engineering Science; Hara, M. [Univ. of Toyama (Japan). Hydrogen Isotope Center; Kobayashi, M. [Shizuoka Univ. (Japan). Radioscience Research Lab.; Oya, Y. [Shizuoka Univ. (Japan). Radioscience Research Lab.; Hatano, Y. [Shizuoka Univ. (Japan). Radioscience Research Lab.

    2015-01-01T23:59:59.000Z

    Tungsten samples were irradiated by neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory at reactor coolant temperatures of 50–70 ?C to low displacement damage of 0.025 and 0.3 dpa. After cooling down, the HFIR neutronirradiated tungsten samples were exposed to deuterium plasmas in the Tritium Plasma Experiment, Idaho National Laboratory at 100, 200 and 500 ?C twice at the ion fluence of 5×1025 m-2 to reach the total ion fluence of 1×1026 m-2 in order to investigate maximum near-surface (<5µm depth) deuterium concentration increased from 0.5 at% D/W in 0.025 dpa samples to 0.8 at% D/W in 0.3 dpa samples. The large discrepancy between the total retention via thermal desorption spectroscopy and the nearsurface retention via nuclear reaction analysis indicated the deuterium was trapped in bulk (at least 50µm depth for 0.025 dpa and 35µm depth for 0.3 dpa) at 500 ?C cases even in the relatively low ion fluence of 1026 m-2.

  6. Hydrogen Cylinder Storage Array Explosion Evaluations at the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Cook, David Howard [ORNL] [ORNL; Griffin, Frederick P [ORNL] [ORNL; Hyman III, Clifton R [ORNL] [ORNL

    2010-01-01T23:59:59.000Z

    The safety analysis for a recently-installed cold neutron source at the High Flux Isotope Reactor (HFIR) involved evaluation of potential explosion consequences from accidental hydrogen jet releases that could occur from an array of hydrogen cylinders. The scope of the safety analysis involved determination of the release rate of hydrogen, the total quantity of hydrogen assumed to be involved in the explosion, the location of an ignition point or center of the explosion from receptors of interest, and the peak overpressure at the receptors. To evaluate the total quantity of hydrogen involved in the explosion, a 2D model was constructed of the jet concentration and a radial-axial integral over the jet cloud from the centerline to the flammability limit of 4% was used to determine the hydrogen mass to be used as a source term. The location of the point source was chosen as the peak of the jet centerline concentration profile. Consequences were assessed using a combination of three methods for estimating local overpressure as a function of explosion source strength and distance: the Baker-Strehlow method, the TNT-equivalence method, and the TNO method. Results from the explosions were assessed using damage estimates in screening tables for buildings and industrial equipment.

  7. Tritium trapping in silicon carbide in contact with solid breeder under high flux isotope reactor irradiation

    SciTech Connect (OSTI)

    H. Katsui; Y. Katoh; A. Hasegawa; M. Shimada; Y. Hatano; T. Hinoki; S. Nogami; T. Tanaka; S. Nagata; T. Shikama

    2013-11-01T23:59:59.000Z

    The trapping of tritium in silicon carbide (SiC) injected from ceramic breeding materials was examined via tritium measurements using imaging plate (IP) techniques. Monolithic SiC in contact with ternary lithium oxide (lithium titanate and lithium aluminate) as a ceramic breeder was irradiated in the High Flux Isotope Reactor (HFIR) in Oak Ridge, Tennessee, USA. The distribution of photo-stimulated luminescence (PSL) of tritium in SiC was successfully obtained, which separated the contribution of 14C ß-rays to the PSL. The tritium incident from ceramic breeders was retained in the vicinity of the SiC surface even after irradiation at 1073 K over the duration of ~3000 h, while trapping of tritium was not observed in the bulk region. The PSL intensity near the SiC surface in contact with lithium titanate was higher than that obtained with lithium aluminate. The amount of the incident tritium and/or the formation of a Li2SiO3 phase on SiC due to the reaction with lithium aluminate under irradiation likely were responsible for this observation.

  8. Simulated Irradiation of Samples in HFIR for use as Possible Test Materials in the MPEX (Material Plasma Exposure Experiment) Facility

    SciTech Connect (OSTI)

    Ellis, Ronald James [ORNL; Rapp, Juergen [ORNL

    2014-01-01T23:59:59.000Z

    The importance of Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) facility will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. The project presented in this paper involved performing assessments of the induced radioactivity and resulting radiation fields of a variety of potential fusion reactor materials. The scientific code packages MCNP and SCALE were used to simulate irradiation of the samples in HFIR; generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. These state-of-the-art simulation methods were used in addressing the challenge of the MPEX project to minimize the radioactive inventory in the preparation of the samples for inclusion in the MPEX facility.

  9. Development of CFD models to support LEU Conversion of ORNL s High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Khane, Vaibhav B [ORNL] [ORNL; Jain, Prashant K [ORNL] [ORNL; Freels, James D [ORNL] [ORNL

    2012-01-01T23:59:59.000Z

    The US Department of Energy s National Nuclear Security Administration (NNSA) is participating in the Global Threat Reduction Initiative to reduce and protect vulnerable nuclear and radiological materials located at civilian sites worldwide. As an integral part of one of NNSA s subprograms, Reduced Enrichment for Research and Test Reactors, HFIR is being converted from the present HEU core to a low enriched uranium (LEU) core with less than 20% of U-235 by weight. Because of HFIR s importance for condensed matter research in the United States, its conversion to a high-density, U-Mo-based, LEU fuel should not significantly impact its existing performance. Furthermore, cost and availability considerations suggest making only minimal changes to the overall HFIR facility. Therefore, the goal of this conversion program is only to substitute LEU for the fuel type in the existing fuel plate design, retaining the same number of fuel plates, with the same physical dimensions, as in the current HFIR HEU core. Because LEU-specific testing and experiments will be limited, COMSOL Multiphysics was chosen to provide the needed simulation capability to validate against the HEU design data and previous calculations, and predict the performance of the proposed LEU fuel for design and safety analyses. To achieve it, advanced COMSOL-based multiphysics simulations, including computational fluid dynamics (CFD), are being developed to capture the turbulent flows and associated heat transfer in fine detail and to improve predictive accuracy [2].

  10. Studies of Plutonium-238 Production at the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Lastres, Oscar [University of Tennessee, Knoxville (UTK)] [University of Tennessee, Knoxville (UTK); Chandler, David [University of Tennessee, Knoxville (UTK) & Oak Ridge National Laboratory (ORNL)] [University of Tennessee, Knoxville (UTK) & Oak Ridge National Laboratory (ORNL); Jarrell, Joshua J [ORNL] [ORNL; Maldonado, G. Ivan [University of Tennessee, Knoxville (UTK)] [University of Tennessee, Knoxville (UTK)

    2011-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2014-10-30T23:59:59.000Z

    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.

  12. HFIR Plant Maintenance - August

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

    Department of Energy FCID Fuel Cycle and Isotopes Division Activities Page 1 of 2 ISO 9001 Comprehensive internal assessment performed Proposal for 100 mA...

  13. A proposed standard on medical isotope production in fission reactors

    SciTech Connect (OSTI)

    Schenter, R. E. [Smart Bullets Inc., 2521 SW Luradel Street, Portland, OR 97219 (United States); Brown, G. J. [Ozarks Medical Center, Cancer Treatment Center, Shaw Medical Building, 1111 Kentucky Avenue, West Plains, MO 65775 (United States); Holden, C. S. [Thorenco LLC, 369 Pine Street, San Francisco, CA 94104 (United States)

    2006-07-01T23:59:59.000Z

    Authors Robert E. Sehenter, Garry Brown and Charles S. Holden argue that a Standard for 'Medical Isotope Production' is needed. Medical isotopes are becoming major components of application for the diagnosis and treatment of all the major diseases including all forms of cancer, heart disease, arthritis, Alzheimer's, among others. Current nuclear data to perform calculations is incomplete, dated or imprecise or otherwise flawed for many isotopes that could have significant applications in medicine. Improved data files will assist computational analyses to design means and methods for improved isotope production techniques in the fission reactor systems. Initial focus of the Standard is expected to be on neutron cross section and branching data for both fast and thermal reactor systems. Evaluated and reviewed tables giving thermal capture cross sections and resonance integrals for the major target and product medical isotopes would be the expected 'first start' for the 'Standard Working Group'. (authors)

  14. HFIR Plant Maintenance - August

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

    IV International Forum working group that is developing design criteria for sodium fast reactors. The meeting was held July 17-18 in Paris, France. * Bob Grove was an invited...

  15. Nuclear reactor fissile isotopes antineutrino spectra

    E-Print Network [OSTI]

    V. Sinev

    2012-07-30T23:59:59.000Z

    Positron spectrum from inverse beta decay reaction on proton was measured in 1988-1990 as a result of neutrino exploration experiment. The measured spectrum has the largest statistics and lowest energy threshold between other neutrino experiments made that time at nuclear reactors. On base of the positron spectrum the standard antineutrino spectrum for typical reactor fuel composition was restored. In presented analysis the partial spectra forming this standard spectrum were extracted using specific method. They could be used for neutrino experiments data analysis made at any fuel composition of reactor core.

  16. CRAD, Maintenance- Oak Ridge National Laboratory High Flux Isotope Reactor

    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 February 2007 assessment of the Maintenance Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  17. CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor

    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 February 2007 assessment of the Management in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  18. CRAD, Training- Oak Ridge National Laboratory High Flux Isotope Reactor

    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 February 2007 assessment of the Training Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  19. CRAD, Engineering- Oak Ridge National Laboratory High Flux Isotope Reactor

    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 February 2007 assessment of the Engineering Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  20. An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

    SciTech Connect (OSTI)

    Rosenthal, Murray Wilford [ORNL

    2009-08-01T23:59:59.000Z

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

  1. Radiation effects on reactor pressure vessel supports

    SciTech Connect (OSTI)

    Johnson, R.E. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology; Lipinski, R.E. [Idaho National Engineering Lab., Rockville, MD (United States)

    1996-05-01T23:59:59.000Z

    The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue.

  2. The use of reduced-moderation light water reactors for transuranic isotope burning in thorium fuel

    E-Print Network [OSTI]

    Lindley, Benjamin A.

    2015-02-03T23:59:59.000Z

    THE USE OF REDUCED-MODERATION LIGHT WATER REACTORS FOR TRANSURANIC ISOTOPE BURNING IN THORIUM FUEL Benjamin Andrew Lindley St Catharine?s College Department of Engineering University of Cambridge A thesis... of Engineering as stated in the Memorandum to Graduate Students. Benjamin Andrew Lindley The Use of Reduced-moderation Light Water Reactors for Transuranic Isotope Burning in Thorium Fuel B. A. Lindley Light water reactors (LWRs) are the world...

  3. Measuring of fissile isotopes partial antineutrino spectra in direct experiment at nuclear reactor

    E-Print Network [OSTI]

    V. V. Sinev

    2009-02-22T23:59:59.000Z

    The direct measuring method is considered to get nuclear reactor antineutrino spectrum. We suppose to isolate partial spectra of the fissile isotopes by using the method of antineutrino spectrum extraction from the inverse beta decay positron spectrum applied at Rovno experiment. This admits to increase the accuracy of partial antineutrino spectra forming the total nuclear reactor spectrum. It is important for the analysis of the reactor core fuel composition and could be applied for non-proliferation purposes.

  4. Type B investigation of the iridium contamination event at the High Flux Isotope Reactor on September 7, 1993

    SciTech Connect (OSTI)

    Not Available

    1994-03-01T23:59:59.000Z

    On the title date, at ORNL, area radiation alarms sounded during a routine transfer of a shielding cask (containing 60 Ci{sup 192}Ir) from the HFIR pool side to a transport truck. Small amounts of Ir were released from the cask onto the reactor bay floor. The floor was cleaned, and the cask was shipped to a hot cell at Building 3047 on Oct. 3, 1993. The event was caused by rupture of one of the Ir target rods after it was loaded into the cask for normal transport operations; the rupture was the result of steam generation in the target rod soon after it was placed in the cask (water had entered the target rod through a tiny defect in a weld while it was in the reactor under pressure). While the target rods were in the reactor and reactor pool, there was sufficient cooling to prevent steam generation; when the target rod was loaded into the dry transport cask, the temperature increased enough to result in boiling of the trapped water and produced high enough pressure to result in rupture. The escaping steam ejected some of the Ir pellets. The event was reported as Occurrence Report Number ORO--MMES-X10HFIR-1993-0030, dated Sept. 8, 1993. Analysis indicated that the following conditions were probable causes: less than adequate welding procedures, practices, or techniques, material controls, or inspection methods, or combination thereof, could have led to weld defects, affecting the integrity of target rod IR-75; less than adequate secondary containment in the cask allowed Ir pellets to escape.

  5. Defect annealing and thermal desorption of deuterium in low dose HFIR neutron-irradiated tungsten

    SciTech Connect (OSTI)

    Masashi Shimada; M. Hara; T. Otsuka; Y. Oya; Y. Hatano

    2014-05-01T23:59:59.000Z

    Accurately estimating tritium retention in plasma facing components (PFCs) and minimizing its uncertainty are key safety issues for licensing future fusion power reactors. D-T fusion reactions produce 14.1 MeV neutrons that activate PFCs and create radiation defects throughout the bulk of the material of these components. Recent studies show that tritium migrates and is trapped in bulk (>> 10 µm) tungsten beyond the detection range of nuclear reaction analysis technique [1-2], and thermal desorption spectroscopy (TDS) technique becomes the only established diagnostic that can reveal hydrogen isotope behavior in in bulk (>> 10 µm) tungsten. Radiation damage and its recovery mechanisms in neutron-irradiated tungsten are still poorly understood, and neutron-irradiation data of tungsten is very limited. In this paper, systematic investigations with repeated plasma exposures and thermal desorption are performed to study defect annealing and thermal desorption of deuterium in low dose neutron-irradiated tungsten. Three tungsten samples (99.99 at. % purity from A.L.M.T. Co., Japan) irradiated at High Flux Isotope Reactor at Oak Ridge National Laboratory were exposed to high flux (ion flux of (0.5-1.0)x1022 m-2s-1 and ion fluence of 1x1026 m-2) deuterium plasma at three different temperatures (100, 200, and 500 °C) in Tritium Plasma Experiment at Idaho National Laboratory. Subsequently, thermal desorption spectroscopy (TDS) was performed with a ramp rate of 10 °C/min up to 900 °C, and the samples were annealed at 900 °C for 0.5 hour. These procedures were repeated three (for 100 and 200 °C samples) and four (for 500 °C sample) times to uncover damage recovery mechanisms and its effects on deuterium behavior. The results show that deuterium retention decreases approximately 90, 75, and 66 % for 100, 200, and 500 °C, respectively after each annealing. When subjected to the same TDS recipe, the desorption temperature shifts from 800 °C to 600 °C after 1st annealing for the sample exposed to TPE at 500 °C. Tritium Migration Analysis Program (TMAP) analysis reveals that the detrapping energy decreases from 1.8 eV to 1.4 eV, indicating the changes in trapping mechanisms. This paper also summarizes deuterium behavior studies in HFIR neutron-irradiated tungsten under US-Japan TITAN program.

  6. Packed bed reactor for photochemical .sup.196 Hg isotope separation

    DOE Patents [OSTI]

    Grossman, Mark W. (Belmont, MA); Speer, Richard (Reading, MA)

    1992-01-01T23:59:59.000Z

    Straight tubes and randomly oriented pieces of tubing having been employed in a photochemical mercury enrichment reactor and have been found to improve the enrichment factor (E) and utilization (U) compared to a non-packed reactor. One preferred embodiment of this system uses a moving bed (via gravity) for random packing.

  7. CRAD, Radiological Controls- Oak Ridge National Laboratory High Flux Isotope Reactor

    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 February 2007 assessment of the Radiation Protection Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  8. CRAD, Training- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR

    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 February 2007 assessment of the Training Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  9. CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR

    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 February 2007 assessment of the Management portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  10. CRAD, Conduct of Operations- Oak Ridge National Laboratory High Flux Isotope Reactor

    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 February, 2007 assessment of the Conduct of Operations Program in preparation for restart of the Oak Ridge National Laboratory, High Flux Isotope Reactor.

  11. CRAD, Engineering- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR

    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 February 2007 assessment of the Engineering Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  12. CRAD, Environmental Protection- Oak Ridge National Laboratory High Flux Isotope Reactor

    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 February 2007 assessment of the Environmental Compliance Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  13. CRAD, Configuration Management- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR

    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 February 2007 assessment of the Configuration Management Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory, High Flux Isotope Reactor.

  14. CRAD, Emergency Management- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR

    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 February 2007 assessment of the Emergency Management Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  15. CRAD, Quality Assurance- Oak Ridge National Laboratory High Flux Isotope Reactor

    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 February 2007 assessment of the Quality Assurance Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  16. CRAD, Safety Basis- Oak Ridge National Laboratory High Flux Isotope Reactor

    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 February 2007 assessment of the Safety Basis in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  17. CRAD, Maintenance- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR

    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 February 2007 assessment of the Maintenance Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  18. CRAD, Conduct of Operations- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR

    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 February, 2007 assessment of the Conduct of Operations Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory, High Flux Isotope Reactor.

  19. CRAD, Occupational Safety & Health- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR

    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 February 2007 assessment of the Occupational Safety and Health Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  20. CRAD, Nuclear Safety- Oak Ridge National Laboratory High Flux Isotope Reactor

    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 February 2007 assessment of the Nuclear Safety Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  1. CRAD, Occupational Safety & Health- Oak Ridge National Laboratory High Flux Isotope Reactor

    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 February 2007 assessment of the Industrial Safety and Hygiene Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  2. CRAD, Safety Basis- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR

    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 February 2007 assessment of the Safety Basis portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  3. CRAD, Configuration Management- Oak Ridge National Laboratory High Flux Isotope Reactor

    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 February 2007 assessment of the Configuration Management Program in preparation for restart of the Oak Ridge National Laboratory, High Flux Isotope Reactor.

  4. CRAD, Emergency Management- Oak Ridge National Laboratory High Flux Isotope Reactor

    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 February 2007 assessment of the Emergency Management Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  5. La138/139 Isotopic Data and Neutron Fluences for Oklo RZ10 Reactor

    E-Print Network [OSTI]

    C. R. Gould; E. I. Sharapov

    2012-08-23T23:59:59.000Z

    Recent years have seen a renewed interest in the Oklo phenomenon, particularly in relation to the study of time variation of the fine structure constant. The neutron fluence is one of the crucial parameters for Oklo reactors. Several approaches to its determination were elaborated in the past. We consider whether it possible to use the present isotopic La138/139 data for RZ10 as an additional indicator of neutron fluences in the active cores of the reactors. We calculate the dependence of the Oklo La138 abundance on neutron fluence and elemental lanthanum concentration. The neutron fluence in RZ10 can be deduced from lanthanum isotopic data, but requires reliable data on the primordial elemental abundance. Conversely, if the fluence is known, the isotope ratio provides information on the primordial lanthanum abundance that is not otherwise easily determined.

  6. Homogeneous fast-flux isotope-production reactor

    DOE Patents [OSTI]

    Cawley, W.E.; Omberg, R.P.

    1982-08-19T23:59:59.000Z

    A method is described for producing tritium in a liquid metal fast breeder reactor. Lithium target material is dissolved in the liquid metal coolant in order to facilitate the production and removal of tritium.

  7. Production capabilities in US nuclear reactors for medical radioisotopes

    SciTech Connect (OSTI)

    Mirzadeh, S.; Callahan, A.P.; Knapp, F.F. Jr. [Oak Ridge National Lab., TN (United States); Schenter, R.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1992-11-01T23:59:59.000Z

    The availability of reactor-produced radioisotopes in the United States for use in medical research and nuclear medicine has traditionally depended on facilities which are an integral part of the US national laboratories and a few reactors at universities. One exception is the reactor in Sterling Forest, New York, originally operated as part of the Cintichem (Union Carbide) system, which is currently in the process of permanent shutdown. Since there are no industry-run reactors in the US, the national laboratories and universities thus play a critical role in providing reactor-produced radioisotopes for medical research and clinical use. The goal of this survey is to provide a comprehensive summary of these production capabilities. With the temporary shutdown of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) in November 1986, the radioisotopes required for DOE-supported radionuclide generators were made available at the Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR). In March 1988, however, the HFBR was temporarily shut down which forced investigators to look at other reactors for production of the radioisotopes. During this period the Missouri University Research Reactor (MURR) played an important role in providing these services. The HFIR resumed routine operation in July 1990 at 85 MW power, and the HFBR resumed operation in June 1991, at 30 MW power. At the time of the HFBR shutdown, there was no available comprehensive overview which could provide information on status of the reactors operating in the US and their capabilities for radioisotope production. The obvious need for a useful overview was thus the impetus for preparing this survey, which would provide an up-to-date summary of those reactors available in the US at both the DOE-funded national laboratories and at US universities where service irradiations are currently or expected to be conducted.

  8. Evolution of isotopic composition of reprocessed uranium during the multiple recycling in light water reactors with natural uranium feed

    SciTech Connect (OSTI)

    Smirnov, A. Yu., E-mail: a.y.smirnoff@rambler.ru; Sulaberidze, G. A. [National Research Nuclear University MEPhI (Russian Federation); Alekseev, P. N.; Dudnikov, A. A.; Nevinitsa, V. A., E-mail: neva@dhtp.kiae.ru; Proselkov, V. N.; Chibinyaev, A. V. [Russian Research Centre Kurchatov Institute (Russian Federation)

    2012-12-15T23:59:59.000Z

    A complex approach based on the consistent modeling of neutron-physics processes and processes of cascade separation of isotopes is applied for analyzing physical problems of the multiple usage of reprocessed uranium in the fuel cycle of light water reactors. A number of scenarios of multiple recycling of reprocessed uranium in light water reactors are considered. In the process, an excess absorption of neutrons by the {sup 236}U isotope is compensated by re-enrichment in the {sup 235}U isotope. Specific consumptions of natural uranium for re-enrichment of the reprocessed uranium depending on the content of the {sup 232}U isotope are obtained.

  9. Monte Carlo uncertainty reliability and isotope production calculations for a fast reactor

    SciTech Connect (OSTI)

    Miles, T.L.

    1992-01-01T23:59:59.000Z

    Statistical uncertainties in Monte Carlo calculations are typically determined by the first and second moments of the tally. For certain types of calculations, there is concern that the uncertainty estimate is significantly non-conservative. This is typically seen in reactor eigenvalue problems where the uncertainty estimate is aggravated by the generation-to-generation fission source. It has been speculated that optimization of the random walk, through biasing techniques, may increase the non-conservative nature of the uncertainty estimate. A series of calculations are documented here which quantify the reliability of the Monte Carlo Neutron and Photon (MCNP) mean and uncertainty estimates by comparing these estimates to the true mean. These calculations were made with a liquid metal fast reactor model, but every effort was made to isolate the statistical nature of the uncertainty estimates so that the analysis of the reliability of the MCNP estimates should be relevant for small thermal reactors as well. Also, preliminary reactor physics calculations for two different special isotope production test assemblies for irradiation in the Fast Flux Test Facility (FFTF) were performed using MCNP and are documented here. The effect of an yttrium-hydride moderator to tailor the neutron flux incident on the targets to maximize isotope production for different designs in different locations within the reactor is discussed. These calculations also demonstrate the useful application of MCNP in design iterations by utilizing many of the codes features.

  10. The DOS 1 neutron dosimetry experiment at the HB-4-A key 7 surveillance site on the HFIR pressure vessel

    SciTech Connect (OSTI)

    Farrell, K.; Kam, F.B.; Baldwin, C.A. [and others

    1994-01-01T23:59:59.000Z

    A comprehensive neutron dosimetry experiment was made at one of the prime surveillance sites at the High Flux Isotope Reactor (HFIR) pressure vessel to aid radiation embrittlement studies of the vessel and to benchmark neutron transport calculations. The thermal neutron flux at the key 7, position 5 site was found, from measurements of radioactivation of four cobalt wires and four silver wires, to be 2.4 {times} 10{sup 12} n{center_dot}m{sup {minus}2}{center_dot}s{sup {minus}1}. The thermal flux derived from two helium accumulation monitors was 2.3 {times} 10{sup 12} n{center_dot}m{sup {minus}2}{center_dot}{sup {minus}1}. The thermal flux estimated by neutron transport calculations was 3.7 {times} 10{sup 12} n{center_dot}m{sup {minus}2}s{sup {minus}1}. The fast flux, >1 MeV, determined from two nickel activation wires, was 1.5 {times} 10{sup 12} n{center_dot}m{sup {minus}2}{center_dot}s{sup {minus}1}, in keeping with values obtained earlier from stainless steel surveillance monitors and with a computed value of 1.2 {times} 10{sup 13} n{center_dot}m{sup {minus}2}{center_dot}{sup {minus}1}. The fast fluxes given by two reaction-product-type monitors, neptunium-237 and beryllium, were 2.6 {times} 10{sup 13} n{center_dot}m{sup {minus}2}{center_dot}s {sup {minus}1} and 2.2 {times} 10{sup 13} n{center_dot}m{sup {minus}2}s{sup {minus}1}, respectively. Follow-up experiments indicate that these latter high values of fast flux are reproducible but are false; they are due to the creation of greater levels of reaction products by photonuclear events induced by an exceptionally high ratio of gamma flux to fast neutron flux at the vessel.

  11. 55Fe effect on enhancing ferritic steel He/dpa ratio in fission reactor irradiations to simulate fusion conditions

    SciTech Connect (OSTI)

    Liu, Haibo; Abdou, Mohamed A.; Greenwood, Lawrence R.

    2013-11-01T23:59:59.000Z

    How to increase the ferritic steel He(appm)/dpa ratio in a fission reactor neutron spectrum is an important question for fusion reactor material testing. An early experiment showed that the accelerated He(appm)/dpa ratio of about 2.3 was achieved for 96% enriched 54Fe in iron with 458.2 effective full power days (EFPD) irradiation in the High Flux Isotope Reactor (HFIR), ORNL. Greenwood suggested that the transmutation produced 55Fe has a thermal neutron helium production cross section which may have an effect on this result. In the current work, the ferritic steel He(appm)/dpa ratio is studied in the neutron spectrum of HFIR with 55Fe thermal neutron helium production taken into account. The available ENDF-b format 55Fe incident neutron cross section file from TENDL, Netherlands, is first input into the calculation model. A benchmark calculation for the same sample as used in the aforementioned experiment was used to adjust and evaluate the TENDL 55Fe (n, a) cross section values. The analysis shows a decrease of a factor of 6700 for the TENDL 55Fe (n, a) cross section in the intermediate and low energy regions is required in order to fit the experimental results. The best fit to the cross section value at thermal neutron energy is about 27 mb. With the adjusted 55Fe (n, a) cross sections, calculation show that the 54Fe and 55Fe isotopes can be enriched by the isotopic tailoring technique in a ferritic steel sample irradiated in HFIR to significantly enhance the helium production rate. The results show that a 70% enriched 54Fe and 30% enriched 55Fe ferritic steel sample would produce a He(appm)/dpa ratio of about 13 initially in the HFIR peripheral target position (PTP). After one year irradiation, the ratio decreases to about 10. This new calculation can be used to guide future isotopic tailoring experiments designed to increase the He(appm)/dpa ratio in fission reactors. A benchmark experiment is suggested to be performed to evaluate the 55Fe (n, a) cross section at thermal energy.

  12. ISOTOPES

    E-Print Network [OSTI]

    Lederer, C. Michael

    2013-01-01T23:59:59.000Z

    §fissile materials in fast breeder reactors currently underwater reactor, FBR = fast breeder reactor. The band belowinc 1 udes heavy-water reactors, fast breeders, and 11

  13. Fuel and core testing plan for a target fueled isotope production reactor.

    SciTech Connect (OSTI)

    Coats, Richard Lee; Dahl, James J.; Parma, Edward J., Jr.

    2010-12-01T23:59:59.000Z

    In recent years there has been an unstable supply of the critical diagnostic medical isotope 99Tc. Several concepts and designs have been proposed to produce 99Mo the parent nuclide of 99Tc, at a commercial scale sufficient to stabilize the world supply. This work lays out a testing and experiment plan for a proposed 2 MW open pool reactor fueled by Low Enriched Uranium (LEU) 99Mo targets. The experiments and tests necessary to support licensing of the reactor design are described and how these experiments and tests will help establish the safe operating envelop for a medical isotope production reactor is discussed. The experiments and tests will facilitate a focused and efficient licensing process in order to bring on line a needed production reactor dedicated to supplying medical isotopes. The Target Fuel Isotope Reactor (TFIR) design calls for an active core region that is approximately 40 cm in diameter and 40 cm in fuel height. It contains up to 150 cylindrical, 1-cm diameter, LEU oxide fuel pins clad with Zircaloy (zirconium alloy), in an annular hexagonal array on a {approx}2.0 cm pitch surrounded, radially, by a graphite or a Be reflector. The reactor is similar to U.S. university reactors in power, hardware, and safety/control systems. Fuel/target pin fabrication is based on existing light water reactor fuel fabrication processes. However, as part of licensing process, experiments must be conducted to confirm analytical predictions of steady-state power and accident conditions. The experiment and test plan will be conducted in phases and will utilize existing facilities at the U.S. Department of Energy's Sandia National Laboratories. The first phase is to validate the predicted reactor core neutronics at delayed critical, zero power and very low power. This will be accomplished by using the Sandia Critical Experiment (CX) platform. A full scale TFIR core will be built in the CX and delayed critical measurements will be taken. For low power experiments, fuel pins can be removed after the experiment and using Sandia's metrology lab, relative power profiles (radially and axially) can be determined. In addition to validating neutronic analyses, confirming heat transfer properties of the target/fuel pins and core will be conducted. Fuel/target pin power limits can be verified with out-of-pile (electrical heating) thermal-hydraulic experiments. This will yield data on the heat flux across the Zircaloy clad and establish safety margin and operating limits. Using Sandia's Annular Core Research Reactor (ACRR) a 4 MW TRIGA type research reactor, target/fuel pins can be driven to desired fission power levels for long durations. Post experiment inspection of the pins can be conducted in the Auxiliary Hot Cell Facility to observe changes in the mechanical properties of the LEU matrix and burn-up effects. Transient tests can also be conducted at the ACRR to observe target/fuel pin performance during accident conditions. Target/fuel pins will be placed in double experiment containment and driven by pulsing the ACRR until target/fuel failure is observed. This will allow for extrapolation of analytical work to confirm safety margins.

  14. Hydrogen loops in existing reactors for testing fuel elements for nuclear propulsion

    SciTech Connect (OSTI)

    Olsen, C.S.; Welland, H.; Abraschoff, J. (Idaho National Engineering Laboratory, EG G Idaho Inc., P.O. Box 1625, Idaho Falls, Idaho 83415 (United States)); Thoms, K. (Oak Ridge National Laboratory, P.O. Box, Oak Ridge, Tennessee 37831-8087 (United States))

    1993-01-15T23:59:59.000Z

    The Space Exploration Initiative (SEI) has revitalized interest in adapting nuclear energy for power and propulsion. Prior to the selection of a nuclear thermal propulsion (NTP) system, extensive testing of the various proposed concepts will be required. In today's environmental, safety and health culture, full size rocket engine tests as were done under the Rover/NERVA program will be extremely difficult and expensive to perform and meet NASA's schedules. A different test strategy uses a hydrogen loop in an existing reactor to test a wide variety of single elements or clusters of elements for fuel qualification. This approach is expected to reduce operating and capital costs and expedite the testing schedule. This paper examines the potential of performing subscale tests in a hydrogen loop in an existing reactor such as the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory or the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. The HFIR is expected to achieve power densities comparable to those achieved in ATR because of the 85 MWt power level and the high thermal and fast flux levels. The available length and diameter of the test region of FHIR are 60 cm and 10 cm whereas the available length and diameter of the test region of ATR are 120 cm and 12 cm respectively.

  15. High Flux Isotope Reactor named Nuclear Historic Landmark | ornl.gov

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville PowerCherries 82981-1cn SunnybankD.jpgHanfordDepartmentInnovationHigh Flux Isotope Reactor named

  16. ISOTOPES

    E-Print Network [OSTI]

    Lederer, C. Michael

    2013-01-01T23:59:59.000Z

    rare (0.017%) isotope 36s at enrichments of 70% at a price32). The enrichment of carbon isotopes by C02-carbamatesulfur isotopes by S02-NaHS03 exchange and the enrichment of

  17. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Gauld, Ian C [ORNL

    2011-01-01T23:59:59.000Z

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  18. Design, construction, and operation of a laboratory scale reactor for the production of high-purity, isotopically enriched bulk silicon

    E-Print Network [OSTI]

    Ager III, J.W.; Beeman, J.W.; Hansen, W.L.; Haller, E.E.

    2004-01-01T23:59:59.000Z

    Russia. The stated isotope enrichments are summarized inenrichments >99% have been achieved for each isotope and

  19. Continuous production of tritium in an isotope-production reactor with a separate circulation system

    DOE Patents [OSTI]

    Cawley, W.E.; Omberg, R.P.

    1982-08-19T23:59:59.000Z

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium is allowed to flow through the reactor in separate loops in order to facilitate the production and removal of tritium.

  20. Isotopes

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville PowerCherries 82981-1cnHigh SchoolIn12 Investigation Peer ReviewIronNuclear Physics » Isotopes

  1. ORNL - Restart of the High Flux Isotope Reactor 2-07

    Broader source: Energy.gov (indexed) [DOE]

    for employees to report safety issues and, when necessary, suspend operations andor stop work. Interviews: Interview selected CS and reactor operators, management, and support...

  2. ORNL - Restart of the High Flux Isotope Reactor 2-07

    Broader source: Energy.gov (indexed) [DOE]

    MAINTENANCE OBJECTIVE MT-1: The maintenance and test programs have been appropriately modified to reflect the CS modification and its reactor interface, sufficient numbers of...

  3. ORNL - Restart of the High Flux Isotope Reactor 2-07

    Broader source: Energy.gov (indexed) [DOE]

    ENGINEERING (ENG) OBJECTIVE ENG-1: The engineering program has been appropriately modified to reflect the CS modification and its reactor interface, sufficient numbers of qualified...

  4. ORNL - Restart of the High Flux Isotope Reactor 2-07

    Broader source: Energy.gov (indexed) [DOE]

    ENGINEERING OBJECTIVE ES-1: The engineering program has been appropriately modified to reflect the CS modification and its reactor interface, sufficient numbers of qualified...

  5. Assemblies with both target and fuel pins in an isotope-production reactor

    DOE Patents [OSTI]

    Cawley, W.E.; Omberg, R.P.

    1982-08-19T23:59:59.000Z

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

  6. Vented target elements for use in an isotope-production reactor. [LMFBR

    DOE Patents [OSTI]

    Cawley, W.E.; Omberg, R.P.

    1982-08-19T23:59:59.000Z

    A method is described for producing tritium gas in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins equipped with vents, and tritium gas is recovered from the coolant.

  7. Fuel pins with both target and fuel pellets in an isotope-production reactor

    DOE Patents [OSTI]

    Cawley, W.E.; Omberg, R.P.

    1982-08-19T23:59:59.000Z

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target pellets are placed in close contact with fissile fuel pellets in order to increase the tritium production rate.

  8. Spallation Neutron Source (SNS) | U.S. DOE Office of Science...

    Office of Science (SC) Website

    (SUF) Division SUF Home About User Facilities User Facilities Dev X-Ray Light Sources Neutron Scattering Facilities High Flux Isotope Reactor (HFIR) Lujan Neutron Scattering...

  9. FY 2013 Summary Report: Post-Irradiation Examination of Zircaloy...

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

    fiscal year 2013 are provided and include information derived from: 1) irradiation of hydrogen-doped zircaloy cladding in High Flux Isotope Reactor (HFIR); 2) mechanical...

  10. CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope...

    Broader source: Energy.gov (indexed) [DOE]

    Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR...

  11. CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope...

    Broader source: Energy.gov (indexed) [DOE]

    Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G...

  12. Neutron Experiment descriptions: N1: Triple-Axis Spectrometers, HFIR HB1A & HB3

    E-Print Network [OSTI]

    Pennycook, Steve

    transport capabilities for large-scale EV applications. Studies on the lithium-ion motion properties phonons in these materials. With the large magnetoelastic interactions in such a material, it is also-Circle Diffractometer, HFIR HB3A Structure and lithium-ion motion in the triphylite LiFePO4 studied by single crystal

  13. Recent Results for the Ferritics Isotopic Tailoring (FIST) Experiment

    SciTech Connect (OSTI)

    Gelles, David S.; Hamilton, M L.; Oliver, Brian M.; Greenwood, Lawrence R.; Ohnuki, Somei; Shiba, K; Kohno, Yutaka; Kohyama, Akira; Robertson, J P.

    2002-12-01T23:59:59.000Z

    An alloy of F82H prepared using the isotope 54 Fe in order to encourage H and He production in a fission reactor has been irradiated in the HFIR JP20 experiment at three temperatures to 7 dpa as TEM disks. Irradiated disks were shear punch tested, examined by TEM, analyzed for He and H content, and compared with previous results in order to quantify irradiation hardening due to transmutation-induced H and He. Hardening due to irradiation is found following irradiation at 300 and 400 C, that is intermediate between that at lower and higher dose, but hardening is negligible following irradiation at 500 C. Microstructural examinations show typical behavior of irradiation as a function of irradiation temperature, with moderate swelling after 400 C irradiation but few bubbles after irradiation at 300 C. Correlations of change in hardening with He and H content show little indication of transmutation-induced hardening, but measured H levels do not agree with predictions and therefore H production and analysis requires further study.

  14. The computerized identification of reactor-produced isotopes in an activation analysis environment

    E-Print Network [OSTI]

    Schlueter, Daniel John

    1971-01-01T23:59:59.000Z

    (ELG ('V rG:", ) I ~ 22F 06(3) 657 80(1000) 1 884 ~ 50( 539) 1 706 80( 162) 1 1384. 30( LO8) 1 686 F 80( 64)1 817, on( 61)1 556. CC( 34) 4 744. 20( 38) 1 937. 30( 243) I 677. 50( 107) 1 1505. ?0( 55)1 1475. 90( 16!1 763. 80( 446. 20( 620. 10( 1562... the rerluirement for U&c d gree of MASTI R OF SCIENC1. December 1971 Major Subject: Computing Science THE COMPUTERIZED IDENTIFICATION OF REACTOR-PRODUCED PSOTOPES IN AN ACTIVATION ANALYSIS ENVIRONMENT A Thesis by DANIEL ZOHN SCIILUETER Approved...

  15. Horizontal Beam Tubes - HFIR Technical Parameters | ORNL Neutron...

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

    Beam Tubes The reactor has four horizontal beam tubes that supply the neutrons to the neutron scattering instruments. Details for each beam tube and instrument can be found on...

  16. Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

    SciTech Connect (OSTI)

    Knight, R.W.; Morin, R.A.

    1999-12-01T23:59:59.000Z

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U{sub 3}O{sub 8} powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated.

  17. Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

    SciTech Connect (OSTI)

    Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N. (U. of Cincinnati)

    2006-02-01T23:59:59.000Z

    A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

  18. The consequences of helium production and nickel additions on microstructure development in isotopically tailored ferritic alloys

    SciTech Connect (OSTI)

    Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States)

    1999-10-01T23:59:59.000Z

    A series of alloys have been made adding various isotopes of nickel in order to vary the production of helium during irradiation by a two step nuclear reaction in a mixed spectrum reactor. The alloys use a base composition of Fe-12Cr with an addition of 1.5% nickel, either in the form of {sup 60}Ni which produces no helium, {sup 59}Ni which produces helium at a rate of about 10 appm He/dpa, or natural nickel which provides an intermediate level of helium due to delayed development of {sup 59}Ni. Specimens were irradiated in the HFIR at Oak Ridge, TN to 7.5 dpa at 300 and 400 C. Microstructural examinations indicated that nickel additions promote precipitation in all alloys, but the effect appears to be much stronger at 400 C than at 300 C. There is sufficient dose by 7 dpa (and with 2 appm He) to initiate void swelling in ferritic/martensitic alloys. Little difference was found between response from {sup 59}Ni and natural nickel. Also, helium bubble development for high helium generation conditions appeared to be very different at 300 and 400 C. At 300 C, it appeared that high densities of bubbles formed whereas at 400 C, bubbles could not be identified, possibly because of the complexity of the microstructure, but more likely because helium accumulated at precipitate interfaces.

  19. The consequences of helium production on microstructural development in isotopically tailored ferritic alloys

    SciTech Connect (OSTI)

    Gelles, D.S. [Pacific Northwest Lab., Richland, WA (United States)

    1996-10-01T23:59:59.000Z

    A series of alloys have been made adding various isotopes of nickel in order to vary the production of helium during irradiation by a two step nuclear reaction in a mixed spectrum reactor. The alloys use a base composition of Fe-12Cr with an addition of 1.5% nickel, either in the form of {sup 60}Ni which produces no helium, {sup 59}Ni which produces helium at a rate of about 10 appm He/dpa, or natural nickel ({sup Nat}Ni) which provides an intermediate level of helium due to delayed development of {sup 59}Ni. Specimens were irradiated in the HFIR at Oak Ridge, TN to {approx}7 dpa at 300 and 400{degrees}C. Microstructural examinations indicated that nickel additions promote precipitation in all alloys, but the effect appears to be much stronger at 400{degrees}C than at 300{degrees}C. There is sufficient dose by 7 dpa (and with 2 appm He) to initiate void swelling in ferritic/martensitic alloys. Little difference was found between response from {sup 59}Ni and {sup Nat}Ni. Also, helium bubble development for high helium generation conditions appeared to be very different at 300 and 400{degrees}C. At 300{degrees}C, it appeared that high densities of bubbles formed whereas at 400{degrees}C, bubbles could not be identified, possibly because of the complexity of the microstructure, but more likely because helium accumulated at precipitate interfaces.

  20. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    SciTech Connect (OSTI)

    A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

    2010-06-01T23:59:59.000Z

    2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed a Readiness Assessment in November 2009. The HSIS is a key component of the ATR National Scientific User Facility (NSUF) operated by Battelle Energy Alliance, LLC and is available to a wide variety of university researchers for nuclear fuels and materials experiments as well as medical isotope research and production.

  1. Overview of the US-Japan collaborative investigation on hydrogen isotope retention in neutron-irradiated and ion-damaged tungsten

    SciTech Connect (OSTI)

    Masashi Shimada; Y. Hatano; Y. Oya; T. Oda; M. Hara; G. Cao; M. Kobayashi; M. Sokolov; H. Watanabe; B. Tyburska; Y. Ueda; P. Calderoni

    2011-09-01T23:59:59.000Z

    Plasma-facing components (PFCs) will be exposed to 14 MeV neutrons from deuterium-tritium (D-T) fusion reactions, and tungsten, a candidate PFC for the divertor in ITER, is expected to receive a neutron dose of 0.7 displacement per atom (dpa) by the end of operation in ITER. The effect of neutron-irradiation damage has been mainly simulated using high-energy ion bombardment. While this prior database of results is quite valuable for understanding the behavior of hydrogen isotopes in PFCs, it does not encompass the full range of effects that must be considered in a practical fusion environment due to short penetration depth, damage gradient, high damage rate, and high PKA energy spectrum of the ion bombardment. In addition, neutrons change the elemental composition via transmutations, and create a high radiation environment inside PFCs, which influence the behavior of hydrogen isotope in PFCs, suggesting the utilization of fission reactors is necessary for neutron irradiation. Therefore, the effort to correlate among high-energy ions, fission neutrons, and fusion neutrons is crucial for accurately estimating tritium retention under a neutron-irradiation environment. Under the framework of the US-Japan TITAN program, tungsten samples (99.99 at. % purity from A.L.M.T. Co.) were irradiated by neutron in the High Flux Isotope Reactor (HFIR), ORNL, at 50 and 300C to 0.025, 0.3, and 1.2 dpa, and the investigation of deuterium retention in neutron-irradiation was performed in the INL Tritium Plasma Experiment (TPE), the unique high-flux linear plasma facility that can handle tritium, beryllium and activated materials. This paper reports the recent results from the comparison of ion-damaged tungsten via various ion species (2.8 MeV Fe2+, 20 MeV W2+, and 700 keV H-) with that from neutron-irradiated tungsten to identify the similarities and differences among them.

  2. Multiscale Modeling of Radiation Damage in

    E-Print Network [OSTI]

    Multiscale Modeling of Radiation Damage in Fusion Reactor Materials Brian D. Wirth, R.J. Kurtz-7405-Eng-48. #12;Presentation overview · Introduction to fusion reactor materials and radiation damage. tailor He HFIR isotopic tailor He HFIR target/RB He appmHe displacement damage (dpa) ffuussiioonn

  3. Characterization of the Neutron Detector Upgrade to the GP-SANS and BIO-SANS Instruments at HFIR

    SciTech Connect (OSTI)

    Berry, Kevin D [ORNL; Bailey, Katherine M [ORNL; Beal, Justin D [ORNL; Diawara, Yacouba [ORNL; Funk, Loren L [ORNL; Hicks, J Steve [ORNL; Jones, Amy Black [ORNL; Littrell, Ken [ORNL; Summers, Randy [ORNL; Urban, Volker S [ORNL; Vandergriff, David H [ORNL; Johnson, Nathan [GE Energy Services; Bradley, Brandon [GE Energy Services

    2012-01-01T23:59:59.000Z

    Over the past year, new 1 m x 1 m neutron detectors have been installed at both the General Purpose SANS (GP-SANS) and the Bio-SANS instruments at HFIR, each intended as an upgrade to provide improved high rate capability. This paper presents the results of characterization studies performed in the detector test laboratory, including position resolution, linearity and background, as well as a preliminary look at high count rate performance.

  4. aerated biofilm reactor: Topics by E-print Network

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

    spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle theta13 by reactor...

  5. astra research reactor: Topics by E-print Network

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

    facilities operated by the department include the research reactor DR 3, the Isotope Laboratory 140 Compound cryopump for fusion reactors CERN Preprints Summary: We reconsider an...

  6. HFIR Vessel Maximum Permissible Pressures for Operating Period 26 to 50 EFPY (100 MW)

    SciTech Connect (OSTI)

    Cheverton, R.D.; Inger, J.R.

    1999-01-01T23:59:59.000Z

    Extending the life of the HFIR pressure vessel from 26 to 50 EFPY (100 MW) requires an updated calculation of the maximum permissible pressure for a range in vessel operating temperatures (40-120 F). The maximum permissible pressure is calculated using the equal-potential method, which takes advantage of knowledge gained from periodic hydrostatic proof tests and uses the test conditions (pressure, temperature, and frequency) as input. The maximum permissible pressure decreases with increasing time between hydro tests but is increased each time a test is conducted. The minimum values that occur just prior to a test either increase or decrease with time, depending on the vessel temperature. The minimum value of these minimums is presently specified as the maximum permissible pressure. For three vessel temperatures of particular interest (80, 88, and 110 F) and a nominal time of 3.0 EFPY(100 MVV)between hydro tests, these pressures are 677, 753, and 850 psi. For the lowest temperature of interest (40 F), the maximum permissible pressure is 295 psi.

  7. Forward model calculations for determining isotopic compositions of materials used in a radiological dispersal device

    E-Print Network [OSTI]

    Burk, David Edward

    2005-08-29T23:59:59.000Z

    -standard lattice assemblies. The measured isotopic concentrations from all three of the reactors showed good agreement with the calculated values....

  8. Exceptional tools for studying the structure and dynamics of materials at the molecular level

    E-Print Network [OSTI]

    Isotope Reactor (HFIR) The highest flux reactor-based neutron source for condensed matter research, industry, and national laboratories include basic tutorials on the principles of scattering theory Neutron Source and High Flux Isotope Reactor facilities. Lecture videos and notes on the following topics

  9. Exceptional tools for studying the structure and dynamics of materials at the molecular level

    E-Print Network [OSTI]

    Isotope Reactor (HFIR) The highest flux reactor-based neutron source for condensed matter research research- ers from academia, industry, and national laboratories include basic tutorials on the principles and Oak Ridge's Spallation Neutron Source and High Flux Isotope Reactor facilities. Lecture videos

  10. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    SciTech Connect (OSTI)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01T23:59:59.000Z

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called User’s Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. User’s week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.

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

    SciTech Connect (OSTI)

    Blandinskiy, V. Yu., E-mail: blandinsky@mail.ru [National Research Center Kurchatov Institute (Russian Federation)

    2014-12-15T23:59:59.000Z

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

  12. Determining Reactor Neutrino Flux

    E-Print Network [OSTI]

    Jun Cao

    2012-03-08T23:59:59.000Z

    Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understanding of the correlation of uncertainties is required for $\\theta_{13}$ experiments. Precise determination of reactor neutrino flux will also improve the sensitivity of the non-proliferation monitoring and future reactor experiments. We will discuss the flux calculation and recent progresses.

  13. Online Catalog of Isotope Products from DOE's National Isotope Development Center

    DOE Data Explorer [Office of Scientific and Technical Information (OSTI)]

    The National Isotope Development Center (NIDC) interfaces with the User Community and manages the coordination of isotope production across the facilities and business operations involved in the production, sale, and distribution of isotopes. A virtual center, the NIDC is funded by the Isotope Development and Production for Research and Applications (IDPRA) subprogram of the Office of Nuclear Physics in the U.S. Department of Energy Office of Science. The Isotope subprogram supports the production, and the development of production techniques of radioactive and stable isotopes that are in short supply for research and applications. Isotopes are high-priority commodities of strategic importance for the Nation and are essential for energy, medical, and national security applications and for basic research; a goal of the program is to make critical isotopes more readily available to meet domestic U.S. needs. This subprogram is steward of the Isotope Production Facility (IPF) at Los Alamos National Laboratory (LANL), the Brookhaven Linear Isotope Producer (BLIP) facility at BNL, and hot cell facilities for processing isotopes at ORNL, BNL and LANL. The subprogram also coordinates and supports isotope production at a suite of university, national laboratory, and commercial accelerator and reactor facilities throughout the Nation to promote a reliable supply of domestic isotopes. The National Isotope Development Center (NIDC) at ORNL coordinates isotope production across the many facilities and manages the business operations of the sale and distribution of isotopes.

  14. Solid tags for identifying failed reactor components

    DOE Patents [OSTI]

    Bunch, Wilbur L. (Richland, WA); Schenter, Robert E. (Richland, WA)

    1987-01-01T23:59:59.000Z

    A solid tag material which generates stable detectable, identifiable, and measurable isotopic gases on exposure to a neutron flux to be placed in a nuclear reactor component, particularly a fuel element, in order to identify the reactor component in event of its failure. Several tag materials consisting of salts which generate a multiplicity of gaseous isotopes in predetermined ratios are used to identify different reactor components.

  15. Producing tritium in a homogenous reactor

    DOE Patents [OSTI]

    Cawley, William E. (Richland, WA)

    1985-01-01T23:59:59.000Z

    A method and apparatus are described for the joint production and separation of tritium. Tritium is produced in an aqueous homogenous reactor and heat from the nuclear reaction is used to distill tritium from the lower isotopes of hydrogen.

  16. Chromatographic hydrogen isotope separation

    DOE Patents [OSTI]

    Aldridge, Frederick T. (Livermore, CA)

    1981-01-01T23:59:59.000Z

    Intermetallic compounds with the CaCu.sub.5 type of crystal structure, particularly LaNiCo.sub.4 and CaNi.sub.5, exhibit high separation factors and fast equilibrium times and therefore are useful for packing a chromatographic hydrogen isotope separation colum. The addition of an inert metal to dilute the hydride improves performance of the column. A large scale mutli-stage chromatographic separation process run as a secondary process off a hydrogen feedstream from an industrial plant which uses large volumes of hydrogen can produce large quantities of heavy water at an effective cost for use in heavy water reactors.

  17. Apparatus for isotopic alteration of mercury vapor

    DOE Patents [OSTI]

    Grossman, Mark W. (Belmont, MA); George, William A. (Gloucester, MA); Marcucci, Rudolph V. (Danvers, MA)

    1988-01-01T23:59:59.000Z

    An apparatus for enriching the isotopic Hg content of mercury is provided. The apparatus includes a reactor, a low pressure electric discharge lamp containing a fill including mercury and an inert gas. A filter is arranged concentrically around the lamp. In a preferred embodiment, constant mercury pressure is maintained in the filter by means of a water-cooled tube that depends from it, the tube having a drop of mercury disposed in it. The reactor is arranged around the filter, whereby radiation from said lamp passes through the filter and into said reactor. The lamp, the filter and the reactor are formed of a material which is transparent to ultraviolet light.

  18. Laser Isotope Enrichment for Medical and Industrial Applications

    SciTech Connect (OSTI)

    Leonard Bond

    2006-07-01T23:59:59.000Z

    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".

  19. from Isotope Production Facility

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

    Cancer-fighting treatment gets boost from Isotope Production Facility April 13, 2012 Isotope Production Facility produces cancer-fighting actinium 2:32 Isotope cancer treatment...

  20. Isotopically controlled semiconductors

    E-Print Network [OSTI]

    Haller, Eugene E.

    2006-01-01T23:59:59.000Z

    16 Isotopically Controlled Semiconductors Eugene E. Hallerof isotopically engineered semiconductors; for outstandingisotopically controlled semiconductor crystals. This article

  1. E-Print Network 3.0 - argonne high flux reactor Sample Search...

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

    for: argonne high flux reactor Page: << < 1 2 3 4 5 > >> 1 Thirteenth National School on Neutron and X-ray Scattering Summary: Neutron Source and High Flux Isotope Reactor...

  2. HFIR Plant Maintenance - August

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

    plasma electrons - Effects back on helicon plasma production - Neutral and plasma density control - RF power-to-plasma heat flux efficiency - Effects of plasma and impurity...

  3. HFIR Plant Maintenance - August

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

    Star-CD and NPhase-CMFD. * We present validation of boiling models in Star-CD and Star- CCM+ for DEBORA and PSBT benchmark problems Sensitivity, verification, and validation...

  4. HFIR Plant Maintenance - August

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

    Multiple DeCART pin-resolved transport Neutronics + CFD Multiple LIME-coupled Star- CCM+ & DeCART pin-resolved transport + commercial computational fluid dynamics (CFD)...

  5. HFIR Plant Maintenance - August

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

    irradiation targets, beginning with small single-pellet capsules ( 237 Np oxide pellets) and progressing towards multi-pellet targets to verify both safety calculations as...

  6. HFIR Plant Maintenance - August

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

    Form Testing of the 10 kg HEU Equivalent RSTD was completed. The resulting test report was submitted to the U.S. Department of Transportation (DOT) as an application for a...

  7. HFIR Plant Maintenance - August

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

    Power Shift (CIPS) while cycles 12 and 14 did not. - CRUD results from deposition of corrosion products on the surface of fuel rods - Boron, contained in the coolant, deposits in...

  8. HFIR Plant Maintenance - August

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

    dictated by TJ-II ECRH cut-off limit (1.7x10 19 m-3) - Large pellets for NBI heating phase (<5x10 19 m-3); two neutral beam systems were previously provided by ORNL in this...

  9. HFIR Plant Maintenance - August

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville Power AdministrationField8,Dist.NewofGeothermal HeatonHEP/NERSC/ASCR Requirementspecu-April 2012 2

  10. HFIR Plant Maintenance - August

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville Power AdministrationField8,Dist.NewofGeothermal HeatonHEP/NERSC/ASCR Requirementspecu-April 2012

  11. HFIR Plant Maintenance - August

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville Power AdministrationField8,Dist.NewofGeothermal HeatonHEP/NERSC/ASCR Requirementspecu-April

  12. HFIR Plant Maintenance - August

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville Power AdministrationField8,Dist.NewofGeothermal HeatonHEP/NERSC/ASCR Requirementspecu-AprilAugust

  13. HFIR Plant Maintenance - August

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville Power AdministrationField8,Dist.NewofGeothermal HeatonHEP/NERSC/ASCR

  14. HFIR Plant Maintenance - August

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville Power AdministrationField8,Dist.NewofGeothermal HeatonHEP/NERSC/ASCRJune 2012 2 Managed by

  15. HFIR Plant Maintenance - August

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville Power AdministrationField8,Dist.NewofGeothermal HeatonHEP/NERSC/ASCRJune 2012 2 Managed byNovember

  16. HFIR Plant Maintenance - August

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville Power AdministrationField8,Dist.NewofGeothermal HeatonHEP/NERSC/ASCRJune 2012 2 Managed

  17. HFIR Plant Maintenance - August

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville Power AdministrationField8,Dist.NewofGeothermal HeatonHEP/NERSC/ASCRJune 2012 2 ManagedSeptember

  18. Observation of the Isotopic Evolution of PWR Fuel Using an Antineutrino Detector

    E-Print Network [OSTI]

    Bowden, N S; Dazeley, S; Svoboda, R; Misner, A; Palmer, T

    2008-01-01T23:59:59.000Z

    By operating an antineutrino detector of simple design during several fuel cycles, we have observed long term changes in antineutrino flux that result from the isotopic evolution of a commercial pressurized water reactor. Measurements made with simple antineutrino detectors of this kind offer an alternative means for verifying fissile inventories at reactors, as part of IAEA and other reactor safeguards regimes.

  19. Development of Technical Nuclear Forensics for Spent Research Reactor Fuel

    E-Print Network [OSTI]

    Sternat, Matthew Ryan 1982-

    2012-11-20T23:59:59.000Z

    , an inverse analysis was developed to re-construct the burnup, initial uranium isotopic compositions, and cooling time of a research reactor spent fuel sample. A convergence acceleration technique was used that consisted of an analytical calculation to predict...

  20. ardennes b-1 reactor: Topics by E-print Network

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

    measurement. We find that with a 3 month exposure of a one ton detector the isotope fractions and the thermal reactor power can be determined at a few percent accuracy, which may...

  1. Phoenix: A Reactor Burnup Code With Uncertainty Quantification

    E-Print Network [OSTI]

    Spence, Grant R

    2014-12-15T23:59:59.000Z

    validation analysis confirmed that the simulation parameters produced by PHOENIX using each perturbation method contained differences of less than five percent for a majority of the cases. The outlying instances where a reactor parameter or isotopic...

  2. Computational neutronics analysis of TRIGA reactors during power pulsing

    E-Print Network [OSTI]

    Bean, Malcolm (Malcolm K.)

    2011-01-01T23:59:59.000Z

    Training, Research, Isotopes, General Atomics (TRIGA) reactors have the unique capability of generating high neutron flux environments with the removal of a transient control rod, creating conditions observed in fast fission ...

  3. Solid State Division progress report for period ending September 30, 1990

    SciTech Connect (OSTI)

    Green, P.H.; Hinton, L.W. (eds.)

    1991-03-01T23:59:59.000Z

    This report covers research progress in the Solid State Division from April 1, 1989, to September 30, 1990. During this period, division research programs were significantly enhanced by the restart of the High-Flux Isotope Reactor (HFIR) and by new initiatives in processing and characterization of materials.

  4. Department founded in 1957 Produced over 1000 graduates in the past

    E-Print Network [OSTI]

    Tennessee, University of

    ) · Awards = $7.1 million · Nuclear Fuels and Materials · Nuclear Security · Radiological Sciences and Health Physics · Nuclear I&C, Reliability, and Safety · Nuclear Fuel Cycles · Advanced Modeling and Simulation · 85 MW High Flux Isotope Reactor (HFIR) · Spallation Neutron Source (SNS) Accelerator · Nuclear

  5. Isotope Science and Production

    E-Print Network [OSTI]

    Isotope Science and Production 35 years of experience in isotope production, processing, and applications. Llllll Committed to the safe and reliable production of radioisotopes, products, and services nuclear materials in trucks and cargo containers. Isotopes for Threat Reduction Isotope production at Los

  6. Stable isotope studies

    SciTech Connect (OSTI)

    Ishida, T.

    1992-01-01T23:59:59.000Z

    The research has been in four general areas: (1) correlation of isotope effects with molecular forces and molecular structures, (2) correlation of zero-point energy and its isotope effects with molecular structure and molecular forces, (3) vapor pressure isotope effects, and (4) fractionation of stable isotopes. 73 refs, 38 figs, 29 tabs.

  7. Nuclear Reactor Safeguards and Monitoring with Antineutrino Detectors

    E-Print Network [OSTI]

    Adam Bernstein; Yifang Wang; Giorgio Gratta; Todd West

    2001-08-01T23:59:59.000Z

    Cubic-meter-sized antineutrino detectors can be used to non-intrusively, robustly and automatically monitor and safeguard a wide variety of nuclear reactor types, including power reactors, research reactors, and plutonium production reactors. Since the antineutrino spectra and relative yields of fissioning isotopes depend on the isotopic composition of the core, changes in composition can be observed without ever directly accessing the core itself. Information from a modest-sized antineutrino detector, coupled with the well-understood principles that govern the core's evolution in time, can be used to determine whether the reactor is being operated in an illegitimate way. A group at Sandia is currently constructing a one cubic meter antineutrino detector at the San Onofre reactor site in California to demonstrate these principles.

  8. Laser-assisted isotope separation of tritium

    DOE Patents [OSTI]

    Herman, Irving P. (Castro Valley, CA); Marling, Jack B. (Livermore, CA)

    1983-01-01T23:59:59.000Z

    Methods for laser-assisted isotope separation of tritium, using infrared multiple photon dissociation of tritium-bearing products in the gas phase. One such process involves the steps of (1) catalytic exchange of a deuterium-bearing molecule XYD with tritiated water DTO from sources such as a heavy water fission reactor, to produce the tritium-bearing working molecules XYT and (2) photoselective dissociation of XYT to form a tritium-rich product. By an analogous procedure, tritium is separated from tritium-bearing materials that contain predominately hydrogen such as a light water coolant from fission or fusion reactors.

  9. Reactor Simulation for Antineutrino Experiments using DRAGON and MURE

    E-Print Network [OSTI]

    C. L. Jones; A. Bernstein; J. M. Conrad; Z. Djurcic; M. Fallot; L. Giot; G. Keefer; A. Onillon; L. Winslow

    2012-06-04T23:59:59.000Z

    Rising interest in nuclear reactors as a source of antineutrinos for experiments motivates validated, fast, and accessible simulations to predict reactor fission rates. Here we present results from the DRAGON and MURE simulation codes and compare them to other industry standards for reactor core modeling. We use published data from the Takahama-3 reactor to evaluate the quality of these simulations against the independently measured fuel isotopic composition. The propagation of the uncertainty in the reactor operating parameters to the resulting antineutrino flux predictions is also discussed.

  10. Reference worldwide model for antineutrinos from reactors

    E-Print Network [OSTI]

    Marica Baldoncini; Ivan Callegari; Giovanni Fiorentini; Fabio Mantovani; Barbara Ricci; Virginia Strati; Gerti Xhixha

    2015-02-16T23:59:59.000Z

    Antineutrinos produced at nuclear reactors constitute a severe source of background for the detection of geoneutrinos, which bring to the Earth's surface information about natural radioactivity in the whole planet. In this framework we provide a reference worldwide model for antineutrinos from reactors, in view of reactors operational records yearly published by the International Atomic Energy Agency (IAEA). We evaluate the expected signal from commercial reactors for ongoing (KamLAND and Borexino), planned (SNO+) and proposed (Juno, RENO-50, LENA and Hanohano) experimental sites. Uncertainties related to reactor antineutrino production, propagation and detection processes are estimated using a Monte Carlo based approach, which provides an overall site dependent uncertainty on the signal in the geoneutrino energy window on the order of 3%. We also implement the off-equilibrium correction to the reference reactor spectra associated with the long-lived isotopes and we estimate a 2.4% increase of the unoscillated event rate in the geoneutrino energy window due to the storage of spent nuclear fuels in the cooling pools. We predict that the research reactors contribute to less than 0.2% to the commercial reactor signal in the investigated 14 sites. We perform a multitemporal analysis of the expected reactor signal over a time lapse of 10 years using reactor operational records collected in a comprehensive database published at www.fe.infn.it/antineutrino.

  11. Candidate processes for diluting the {sup 235}U isotope in weapons-capable highly enriched uranium

    SciTech Connect (OSTI)

    Snider, J.D.

    1996-02-01T23:59:59.000Z

    The United States Department of Energy (DOE) is evaluating options for rendering its surplus inventories of highly enriched uranium (HEU) incapable of being used to produce nuclear weapons. Weapons-capable HEU was earlier produced by enriching uranium in the fissile {sup 235}U isotope from its natural occurring 0.71 percent isotopic concentration to at least 20 percent isotopic concentration. Now, by diluting its concentration of the fissile {sup 235}U isotope in a uranium blending process, the weapons capability of HEU can be eliminated in a manner that is reversible only through isotope enrichment, and therefore, highly resistant to proliferation. To the extent that can be economically and technically justified, the down-blended uranium product will be made suitable for use as commercial reactor fuel. Such down-blended uranium product can also be disposed of as waste if chemical or isotopic impurities preclude its use as reactor fuel.

  12. Heterogeneous Recycling in Fast Reactors

    SciTech Connect (OSTI)

    Forget, Benoit; Pope, Michael; Piet, Steven J.; Driscoll, Michael

    2012-07-30T23:59:59.000Z

    Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

  13. Solid State Reactor Final Report

    SciTech Connect (OSTI)

    Mays, G.T.

    2004-03-10T23:59:59.000Z

    The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas of research were undertaken: (1) establishing the design and safety-related basis via neutronic and reactor control assessments with the graphite foam as heat transfer medium; (2) evaluating the thermal performance of the graphite foam for heat removal, reactor stability, reactor operations, and overall core thermal characteristics; (3) characterizing the physical properties of the graphite foam under normal and irradiated conditions to determine any effects on structure, dimensional stability, thermal conductivity, and thermal expansion; and (4) developing a power conversion system design to match the reactor operating parameters.

  14. Nuclear reactor engineering

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1981-01-01T23:59:59.000Z

    Chapters are presented concerning energy from nuclear fission; nuclear reactions and radiations; diffusion and slowing-down of neutrons; principles of reactor analysis; nuclear reactor kinetics and control; energy removal; non-fuel reactor materials; the reactor fuel system; radiation protection and environmental effects; nuclear reactor shielding; nuclear reactor safety; and power reactor systems.

  15. Developing the Sandia National Laboratories transportation infrastructure for isotope products and wastes

    SciTech Connect (OSTI)

    Trennel, A.J.

    1995-12-31T23:59:59.000Z

    Certain radioactive isotopes for North American and especially the United States` needs are enormously important to the medical community and their numerous patients. The most important medical isotope is {sup 99}Mo, which is currently manufactured by Nordion International Inc. in a single, aging reactor operated by Atomic Energy of Canada, Ltd. The reactor`s useful life is expected to end at the turn of the century. Production loss because of reactor shutdown possibilities prompted the US Congress to direct the DOE to provide for a US backup source for this crucial isotope. The SNL Annular Core Research Reactor (ACRR) was evaluated as a site to provide {sup 99}Mo initially and other isotopes that can be economically extracted from the process. Medical isotope production at SNL is a new venture in manufacturing. Should SNL be selected and the project reach the manufacturing stage, SNL would expect to service up to 30% of the US market under normal circumstances as a backup to the Canadian supply with the capability to service 100% should the need arise. The demand for {sup 99}Mo increases each year; hence, the proposed action accommodates growth in demand to meet this increase. The proposed project would guarantee the supply of medical isotopes would continue if either the irradiation or processing activities in Canada were interrupted.

  16. Bioconversion reactor

    DOE Patents [OSTI]

    McCarty, Perry L. (Stanford, CA); Bachmann, Andre (Palo Alto, CA)

    1992-01-01T23:59:59.000Z

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  17. Catalytic reactor

    DOE Patents [OSTI]

    Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

    2009-03-10T23:59:59.000Z

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  18. 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. (Los Alamos National Laboratory, Los Alamos, NM); Smith, James Dean; Longmire, Pamela

    2010-04-01T23:59:59.000Z

    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.

  19. Hybrid isotope separation scheme

    DOE Patents [OSTI]

    Maya, J.

    1991-06-18T23:59:59.000Z

    A method is described for yielding selectively a desired enrichment in a specific isotope including the steps of inputting into a spinning chamber a gas from which a scavenger, radiating the gas with a wave length or frequency characteristic of the absorption of a particular isotope of the atomic or molecular gas, thereby inducing a photochemical reaction between the scavenger, and collecting the specific isotope-containing chemical by using a recombination surface or by a scooping apparatus. 2 figures.

  20. Hybrid isotope separation scheme

    DOE Patents [OSTI]

    Maya, Jakob (Brookline, MA)

    1991-01-01T23:59:59.000Z

    A method of yielding selectively a desired enrichment in a specific isotope including the steps of inputting into a spinning chamber a gas from which a scavenger, radiating the gas with a wave length or frequency characteristic of the absorption of a particular isotope of the atomic or molecular gas, thereby inducing a photochemical reaction between the scavenger, and collecting the specific isotope-containing chemical by using a recombination surface or by a scooping apparatus.

  1. Stable isotope enrichment

    ScienceCinema (OSTI)

    Egle, Brian

    2014-07-15T23:59:59.000Z

    Brian Egle is working to increase the nation's capacity to produce stable isotopes for use including medicine, industry and national security.

  2. Isotopically controlled semiconductors

    E-Print Network [OSTI]

    Haller, E.E.

    2004-01-01T23:59:59.000Z

    and phonons in semiconductors,” J. Non-Cryst. Solids 141 (LVM) Spectroscopy of Semiconductors,” Mat. Res. Soc. Symp.Isotopically Engineered Semiconductors – New Media for the

  3. Stable isotope enrichment

    SciTech Connect (OSTI)

    Egle, Brian

    2014-07-14T23:59:59.000Z

    Brian Egle is working to increase the nation's capacity to produce stable isotopes for use including medicine, industry and national security.

  4. Deciphering the measured ratios of Iodine-131 to Cesium-137 at the Fukushima reactors

    E-Print Network [OSTI]

    Matsui, T

    2011-01-01T23:59:59.000Z

    We calculate the relative abundance of the radioactive isotopes Iodine-131 and Cesium-137 produced by nuclear fission in reactors and compare it with data taken at the troubled Fukushima Dai-ichi nuclear power plant. The ratio of radioactivities of these two isotopes can be used to obtain information about when the nuclear reactions terminated.

  5. Deciphering the measured ratios of Iodine-131 to Cesium-137 at the Fukushima reactors

    E-Print Network [OSTI]

    T. Matsui

    2011-12-13T23:59:59.000Z

    We calculate the relative abundance of the radioactive isotopes Iodine-131 and Cesium-137 produced by nuclear fission in reactors and compare it with data taken at the troubled Fukushima Dai-ichi nuclear power plant. The ratio of radioactivities of these two isotopes can be used to obtain information about when the nuclear reactions terminated.

  6. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J. (Augusta, GA); Babcock, Dale F. (Wilmington, DE); Menegus, Robert L. (Wilmington, DE)

    1983-01-01T23:59:59.000Z

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  7. Discovery of the Tungsten Isotopes

    E-Print Network [OSTI]

    A. Fritsch; J. Q. Ginepro; M. Heim; A. Schuh; A. Shore; M. Thoennessen

    2009-03-25T23:59:59.000Z

    Thirty-five tungsten isotopes have so far been observed; the discovery of these isotopes is discussed. For each isotope a brief summary of the first refereed publication, including the production and identification method, is presented.

  8. Discovery of the Titanium Isotopes

    E-Print Network [OSTI]

    D. Meierfrankenfeld; M. Thoennessen

    2010-09-08T23:59:59.000Z

    Twentyfive titanium isotopes have so far been observed; the discovery of these isotopes is discussed. For each isotope a brief summary of the first refereed publication, including the production and identification method, is presented.

  9. Discovery of the Tin Isotopes

    E-Print Network [OSTI]

    S. Amos; M. Thoennessen

    2010-09-08T23:59:59.000Z

    Thirty-eight tin isotopes have so far been observed; the discovery of these isotopes is discussed. For each isotope a brief summary of the first refereed publication, including the production and identification method, is presented.

  10. Discovery of the tungsten isotopes

    SciTech Connect (OSTI)

    Fritsch, A.; Ginepro, J.Q.; Heim, M.; Schuh, A.; Shore, A. [National Superconducting Cyclotron Laboratory and Department of Physics and Astronomy, Michigan State University, East Lansing, MI 48824 (United States); Thoennessen, M. [National Superconducting Cyclotron Laboratory and Department of Physics and Astronomy, Michigan State University, East Lansing, MI 48824 (United States)], E-mail: thoennessen@nscl.msu.edu

    2010-05-15T23:59:59.000Z

    Thirty-five tungsten isotopes have been observed so far and the discovery of these isotopes is discussed here. For each isotope a brief summary of the first refereed publication, including the production and identification method, is presented.

  11. Discovery of the Tungsten Isotopes

    E-Print Network [OSTI]

    Fritsch, A; Heim, M; Schuh, A; Shore, A; Thoennessen, M

    2009-01-01T23:59:59.000Z

    Thirty-five tungsten isotopes have so far been observed; the discovery of these isotopes is discussed. For each isotope a brief summary of the first refereed publication, including the production and identification method, is presented.

  12. A reference worldwide model for antineutrinos from reactors

    E-Print Network [OSTI]

    Baldoncini, Marica; Fiorentini, Giovanni; Mantovani, Fabio; Ricci, Barbara; Strati, Virginia; Xhixha, Gerti

    2014-01-01T23:59:59.000Z

    Antineutrinos produced at nuclear reactors constitute a severe source of background for the detection of geoneutrinos, which bring to the Earth's surface information about natural radioactivity in the whole planet. In this framework we provide a reference worldwide model for antineutrinos from reactors, in view of reactors operational records yearly published by the International Atomic Energy Agency (IAEA). We evaluate the expected signal from commercial reactors for ongoing (KamLAND and Borexino), planned (SNO+) and proposed (Juno, RENO-50, LENA and Hanohano) experimental sites. Uncertainties related to reactor antineutrino production, propagation and detection processes are estimated using a Monte Carlo based approach, which provides an overall site dependent uncertainty on the signal in the geoneutrino energy window on the order of 3%. We also implement the off-equilibrium correction to the reference reactor spectra associated with the long-lived isotopes and we estimate a 2.4% increase of the unoscillate...

  13. REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

    SciTech Connect (OSTI)

    Nichols, T.; Beals, D.; Sternat, M.

    2011-07-18T23:59:59.000Z

    Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Many research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.

  14. (Carbon isotope fractionation inplants)

    SciTech Connect (OSTI)

    O'Leary, M.H.

    1990-01-01T23:59:59.000Z

    The objectives of this research are: To develop a theoretical and experimental framework for understanding isotope fractionations in plants; and to develop methods for using this isotope fractionation for understanding the dynamics of CO{sub 2} fixation in plants. Progress is described.

  15. Laser isotope separation

    DOE Patents [OSTI]

    Robinson, C. Paul (Los Alamos, NM); Jensen, Reed J. (Los Alamos, NM); Cotter, Theodore P. (Munich, DE); Boyer, Keith (Los Alamos, NM); Greiner, Norman R. (Los Alamos, NM)

    1988-01-01T23:59:59.000Z

    A process and apparatus for separating isotopes by selective excitation of isotopic species of a volatile compound by tuned laser light. A highly cooled gas of the volatile compound is produced in which the isotopic shift is sharpened and defined. Before substantial condensation occurs, the cooled gas is irradiated with laser light precisely tuned to a desired wavelength to selectively excite a particular isotopic species in the cooled gas. The laser light may impart sufficient energy to the excited species to cause it to undergo photolysis, photochemical reaction or even to photoionize. Alternatively, a two-photon irradiation may be applied to the cooled gas to induce photolysis, photochemical reaction or photoionization. The process is particularly applicable to the separation of isotopes of uranium.

  16. Photochemical isotope separation

    DOE Patents [OSTI]

    Robinson, C. Paul (Los Alamos, NM); Jensen, Reed J. (Los Alamos, NM); Cotter, Theodore P. (Los Alamos, NM); Greiner, Norman R. (Los Alamos, NM); Boyer, Keith (Los Alamos, NM)

    1987-01-01T23:59:59.000Z

    A process for separating isotopes by selective excitation of isotopic species of a volatile compound by tuned laser light. A highly cooled gas of the volatile compound is produced in which the isotopic shift is sharpened and defined. Before substantial condensation occurs, the cooled gas is irradiated with laser light precisely tuned to a desired wavelength to selectively excite a particular isotopic species in the cooled gas. The laser light may impart sufficient energy to the excited species to cause it to undergo photochemical reaction or even to photoionize. Alternatively, a two-photon irradiation may be applied to the cooled gas to induce photochemical reaction or photoionization. The process is particularly applicable to the separation of isotopes of uranium and plutonium.

  17. Photochemical isotope separation

    DOE Patents [OSTI]

    Robinson, C.P.; Jensen, R.J.; Cotter, T.P.; Greiner, N.R.; Boyer, K.

    1987-04-28T23:59:59.000Z

    A process is described for separating isotopes by selective excitation of isotopic species of a volatile compound by tuned laser light. A highly cooled gas of the volatile compound is produced in which the isotopic shift is sharpened and defined. Before substantial condensation occurs, the cooled gas is irradiated with laser light precisely tuned to a desired wavelength to selectively excite a particular isotopic species in the cooled gas. The laser light may impart sufficient energy to the excited species to cause it to undergo photochemical reaction or even to photoionize. Alternatively, a two-photon irradiation may be applied to the cooled gas to induce photochemical reaction or photoionization. The process is particularly applicable to the separation of isotopes of uranium and plutonium. 8 figs.

  18. Axi-symmetrical flow reactor for .sup.196 Hg photochemical enrichment

    DOE Patents [OSTI]

    Grossman, Mark W. (Belmont, MA)

    1991-01-01T23:59:59.000Z

    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.

  19. Axi-symmetrical flow reactor for [sup 196]Hg photochemical enrichment

    DOE Patents [OSTI]

    Grossman, M.W.

    1991-04-30T23:59:59.000Z

    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.

  20. Structure of processes in flow reactor and closed reactor: Flow reactor

    E-Print Network [OSTI]

    Greifswald, Ernst-Moritz-Arndt-Universität

    Structure of processes in flow reactor and closed reactor: Flow reactor Closed reactor Active Zone -- chemical quasi- equilibria, similarity principles and macroscopic kinetics", in: Lectures on Plasma Physics

  1. Spectral Structure of Electron Antineutrinos from Nuclear Reactors

    E-Print Network [OSTI]

    D. A. Dwyer; T. J. Langford

    2014-07-04T23:59:59.000Z

    Recent measurements of the positron energy spectrum obtained from inverse beta decay interactions of reactor electron antineutrinos show an excess in the 4 to 6 MeV region relative to current predictions. First-principle calculations of fission and beta decay processes within a typical pressurized water reactor core identify prominent fission daughter isotopes as a possible origin for this excess. These calculations also predict percent-level substructure in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of this substructure can constrain nuclear reactor physics. The substructure can be a systematic uncertainty for measurements utilizing the detailed spectral shape.

  2. Spectral Structure of Electron Antineutrinos from Nuclear Reactors

    E-Print Network [OSTI]

    Dwyer, D A

    2014-01-01T23:59:59.000Z

    Recent measurements of the positron energy spectrum obtained from inverse beta decay interactions of reactor electron antineutrinos show an excess in the 4 to 6 MeV region relative to current predictions. First-principle calculations of fission and beta decay processes within a typical pressurized water reactor core identify prominent fission daughter isotopes as a possible origin for this excess. These calculations also predict percent-level substructure in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of this substructure can constrain nuclear reactor physics. The substructure can be a systematic uncertainty for measurements utilizing the detailed spectral shape.

  3. Tritium Formation and Mitigation in High Temperature Reactors

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots

    2012-08-01T23:59:59.000Z

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. In order to prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for the production of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750°C. Results of the diffusion model are presented for one steadystate value of tritium production in the reactor.

  4. PART II, Tackling Grand Challenges in Geochemistry: Q&A with...

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

    for scientific research and industrial development, and HFIR is a powerful reactor-based source. We use these in a couple of ways. One way (at HFIR) relies on the fact that...

  5. Hybrid adsorptive membrane reactor

    DOE Patents [OSTI]

    Tsotsis, Theodore T. (Huntington Beach, CA); Sahimi, Muhammad (Altadena, CA); Fayyaz-Najafi, Babak (Richmond, CA); Harale, Aadesh (Los Angeles, CA); Park, Byoung-Gi (Yeosu, KR); Liu, Paul K. T. (Lafayette Hill, PA)

    2011-03-01T23:59:59.000Z

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  6. Nuclear reactor engineering

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1982-07-01T23:59:59.000Z

    A book is reviewed which emphasizes topics directly related to the light water reactor power plant and the fast reactor power system. Current real-world problems are addressed throughout the text, and a chapter on safety includes much of the postThree Mile Island impact on operating systems. Topics covered include Doppler broadening, neutron resonances, multigroup diffusion theory, reactor kinetics, reactor control, energy removal, nonfuel materials, reactor fuel, radiation protection, environmental effects, and reactor safety.

  7. High-purity, isotopically enriched bulk silicon

    E-Print Network [OSTI]

    2004-01-01T23:59:59.000Z

    Russia. The stated isotope enrichments are summarized inenrichments >99% have been achieved for each isotope andthe enrichment is highest, are presented. isotope at. % nat.

  8. Hydrogen isotope separation utilizing bulk getters

    DOE Patents [OSTI]

    Knize, Randall J. (Los Angeles, CA); Cecchi, Joseph L. (Lawrenceville, NJ)

    1990-01-01T23:59:59.000Z

    Tritium and deuterium are separated from a gaseous mixture thereof, derived from a nuclear fusion reactor or some other source, by providing a casing with a bulk getter therein for absorbing the gaseous mixture to produce an initial loading of the getter, partially desorbing the getter to produce a desorbed mixture which is tritium-enriched, pumping the desorbed mixture into a separate container, the remaining gaseous loading in the getter being deuterium-enriched, desorbing the getter to a substantially greater extent to produce a deuterium-enriched gaseous mixture, and removing the deuterium-enriched mixture into another container. The bulk getter may comprise a zirconium-aluminum alloy, or a zirconium-vanadium-iron alloy. The partial desorption may reduce the loading by approximately fifty percent. The basic procedure may be extended to produce a multistage isotope separator, including at least one additional bulk getter into which the tritium-enriched mixture is absorbed. The second getter is then partially desorbed to produce a desorbed mixture which is further tritium-enriched. The last-mentioned mixture is then removed from the container for the second getter, which is then desorbed to a substantially greater extent to produce a desorbed mixture which is deuterium-enriched. The last-mentioned mixture is then removed so that the cycle can be continued and repeated. The method of isotope separation is also applicable to other hydrogen isotopes, in that the method can be employed for separating either deuterium or tritium from normal hydrogen.

  9. Hydrogen isotope separation utilizing bulk getters

    DOE Patents [OSTI]

    Knize, Randall J. (Los Angeles, CA); Cecchi, Joseph L. (Lawrenceville, NJ)

    1991-01-01T23:59:59.000Z

    Tritium and deuterium are separated from a gaseous mixture thereof, derived from a nuclear fusion reactor or some other source, by providing a casing with a bulk getter therein for absorbing the gaseous mixture to produce an initial loading of the getter, partially desorbing the getter to produce a desorbed mixture which is tritium-enriched, pumping the desorbed mixture into a separate container, the remaining gaseous loading in the getter being deuterium-enriched, desorbing the getter to a substantially greater extent to produce a deuterium-enriched gaseous mixture, and removing the deuterium-enriched mixture into another container. The bulk getter may comprise a zirconium-aluminum alloy, or a zirconium-vanadium-iron alloy. The partial desorption may reduce the loading by approximately fifty percent. The basic procedure may be extended to produce a multistage isotope separator, including at least one additional bulk getter into which the tritium-enriched mixture is absorbed. The second getter is then partially desorbed to produce a desorbed mixture which is further tritium-enriched. The last-mentioned mixture is then removed from the container for the second getter, which is then desorbed to a substantially greater extent to produce a desorbed mixture which is deuterium-enriched. The last-mentioned mixture is then removed so that the cycle can be continued and repeated. The method of isotope separation is also applicable to other hydrogen isotopes, in that the method can be employed for separating either deuterium or tritium from normal hydrogen.

  10. Hydrogen isotope separation utilizing bulk getters

    DOE Patents [OSTI]

    Knize, R.J.; Cecchi, J.L.

    1991-08-20T23:59:59.000Z

    Tritium and deuterium are separated from a gaseous mixture thereof, derived from a nuclear fusion reactor or some other source, by providing a casing with a bulk getter therein for absorbing the gaseous mixture to produce an initial loading of the getter, partially desorbing the getter to produce a desorbed mixture which is tritium-enriched, pumping the desorbed mixture into a separate container, the remaining gaseous loading in the getter being deuterium-enriched, desorbing the getter to a substantially greater extent to produce a deuterium-enriched gaseous mixture, and removing the deuterium-enriched mixture into another container. The bulk getter may comprise a zirconium-aluminum alloy, or a zirconium-vanadium-iron alloy. The partial desorption may reduce the loading by approximately fifty percent. The basic procedure may be extended to produce a multistage isotope separator, including at least one additional bulk getter into which the tritium-enriched mixture is absorbed. The second getter is then partially desorbed to produce a desorbed mixture which is further tritium-enriched. The last-mentioned mixture is then removed from the container for the second getter, which is then desorbed to a substantially greater extent to produce a desorbed mixture which is deuterium-enriched. The last-mentioned mixture is then removed so that the cycle can be continued and repeated. The method of isotope separation is also applicable to other hydrogen isotopes, in that the method can be employed for separating either deuterium or tritium from normal hydrogen. 4 figures.

  11. A one-group parametric sensitivity analysis for the graphite isotope ratio method and other related techniques using ORIGEN 2.2 

    E-Print Network [OSTI]

    Chesson, Kristin Elaine

    2009-06-02T23:59:59.000Z

    Several methods have been developed previously for estimating cumulative energy production and plutonium production from graphite-moderated reactors. The Graphite Isotope Ratio Method (GIRM) is one well-known technique. This method is based...

  12. Predicting Reactor Antineutrino Emissions Using New Precision Beta Spectroscopy

    SciTech Connect (OSTI)

    Asner, David M.; Burns, Kimberly A.; Greenfield, Bryce A.; Kos, Marek S.; Orrell, John L.; Schram, Malachi; VanDevender, Brent A.; Wootan, David W.

    2013-05-01T23:59:59.000Z

    Neutrino experiments at nuclear reactors are currently vital to the study of neutrino oscillations. The observed antineutrino rates at reactors are typically lower than model expectations. This observed deficit is called the “reactor neutrino anomaly”. A new understanding of neutrino physics may be required to explain this deficit, though model estimation uncertainties may also play a role in the apparent discrepancy. PNNL is currently investigating an experimental technique that promises reduced uncertainties for measured data to support these hypotheses and interpret reactor antineutrino measurements. The experimental approach is to 1) direct a proton accelerator beam on a metal target to produce a source of neutrons, 2) use spectral tailoring to modify the neutron spectrum to closely simulate the energy distribution of a power reactor neutron spectrum, 3) irradiate isotopic fission foils (235U, 238U, 239Pu, 241Pu) in this neutron spectrum so that fissions occur at energies representative of a reactor, 4) transport the beta particles released by the fission products in the foils to a beta spectrometer, 5) measure the beta energy spectrum, and 6) invert the measured beta energy spectrum to an antineutrino energy spectrum. A similar technique using a beta spectrometer and isotopic fission foils was pioneered in the 1980’s at the ILL thermal reactor. Those measurements have been the basis for interpreting all subsequent antineutrino measurements at reactors. A basic constraint in efforts to reduce uncertainties in predicting the antineutrino emission from reactor cores is any underlying limitation of the original measurements. This may include beta spectrum energy resolution, the absolute normalization of beta emission to number of fission, statistical counting uncertainties, lack of 238U data, the purely thermal nature of the IIL reactor neutrons used, etc. An accelerator-based neutron source that can be tailored to match various reactor neutron spectra provides an advantage for control in studying how changes in the neutron spectra (i.e. "in the reactor core") affects the resulting fission product beta spectrum. Furthermore, the 238U antineutrino spectrum, which has not been measured, can be studied directly because of the enhanced 1 MeV fast neutron flux available at the accelerator source. A facility such as the Project X Injector Experiment (PXIE) 30 MeV proton linear accelerator at Fermilab is being considered for this experiment. The hypothesis is that a new approach utilizing the flexibility of an accelerator neutron source with spectral tailoring coupled with a careful design of an isotopic fission target and beta spectrometer and the inversion of the beta spectrum to the neutrino spectrum will allow further reduction in the uncertainties associated with prediction of the reactor antineutrino spectrum.

  13. Transportation of medical isotopes

    SciTech Connect (OSTI)

    Nielsen, D.L.

    1997-11-19T23:59:59.000Z

    A Draft Technical Information Document (HNF-1855) is being prepared to evaluate proposed interim tritium and medical isotope production at the Fast Flux Test Facility (FFTF). This assessment examines the potential health and safety impacts of transportation operations associated with the production of medical isotopes. Incident-free and accidental impacts are assessed using bounding source terms for the shipment of nonradiological target materials to the Hanford Site, the shipment of irradiated targets from the FFTF to the 325 Building, and the shipment of medical isotope products from the 325 Building to medical distributors. The health and safety consequences to workers and the public from the incident-free transportation of targets and isotope products would be within acceptable levels. For transportation accidents, risks to works and the public also would be within acceptable levels. This assessment is based on best information available at this time. As the medical isotope program matures, this analysis will be revised, if necessary, to support development of a final revision to the Technical Information Document.

  14. SRS Small Modular Reactors

    SciTech Connect (OSTI)

    None

    2012-04-27T23:59:59.000Z

    The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

  15. SRS Small Modular Reactors

    ScienceCinema (OSTI)

    None

    2014-05-21T23:59:59.000Z

    The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

  16. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J. (Clairton, PA)

    1980-03-11T23:59:59.000Z

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

  17. Control Rod Malfunction at the NRAD Reactor

    SciTech Connect (OSTI)

    Thomas L. Maddock

    2010-05-01T23:59:59.000Z

    The neutron Radiography Reactor (NRAD) is a training, research, and isotope (TRIGA) reactor located at the INL. The reactor is normally shut down by the insertion of three control rods that drop into the core when power is removed from electromagnets. During a routine shutdown, indicator lights on the console showed that one of the control rods was not inserted. It was initially thought that the indicator lights were in error because of a limit switch that was out of adjustment. Through further testing, it was determined that the control rod did not drop when the scram switch was initially pressed. The control rod anomaly led to a six month shutdown of the reactor and an in depth investigation of the reactor protective system. The investigation looked into: scram switch operation, console modifications, and control rod drive mechanisms. A number of latent issues were discovered and corrected during the investigation. The cause of the control rod malfunction was found to be a buildup of corrosion in the control rod drive mechanism. The investigation resulted in modifications to equipment, changes to both operation and maintenance procedures, and additional training. No reoccurrences of the problem have been observed since corrective actions were implemented.

  18. Isotope Production at the Hanford Site in Richland, Washington

    SciTech Connect (OSTI)

    Ammoniums

    1999-06-01T23:59:59.000Z

    This report was prepared in response to a request from the Nuclear Energy Research Advisory Committee (NERAC) subcommittee on ''Long-Term Isotope Research and Production Plans.'' The NERAC subcommittee has asked for a reply to a number of questions regarding (1) ''How well does the Department of Energy (DOE) infrastructure sme the need for commercial and medical isotopes?'' and (2) ''What should be the long-term role of the federal government in providing commercial and medical isotopes?' Our report addresses the questions raised by the NERAC subcommittee, and especially the 10 issues that were raised under the first of the above questions (see Appendix). These issues are related to the isotope products offered by the DOE Isotope Production Sites, the capabilities and condition of the facilities used to produce these products, the management of the isotope production programs at DOE laboratories, and the customer service record of the DOE Isotope Production sites. An important component of our report is a description of the Fast Flux Test Facility (FFTF) reactor at the Hbford Site and the future plans for its utilization as a source of radioisotopes needed by nuclear medicine physicians, by researchers, and by customers in the commercial sector. In response to the second question raised by the NERAC subcommittee, it is our firm belief that the supply of isotopes provided by DOE for medical, industrial, and research applications must be strengthened in the near future. Many of the radioisotopes currently used for medical diagnosis and therapy of cancer and other diseases are imported from Canada, Europe, and Asia. This situation places the control of isotope availability, quality, and pricing in the hands of non-U.S. suppliers. It is our opinion that the needs of the U.S. customers for isotopes and isotope products are not being adequately served, and that the DOE infrastructure and facilities devoted to the supply of these products must be improved This perception forms one of the fundamental bases for our proposal that the FFTF, which is currently in a standby condition, be reactivated to supply nuclear services and products such as radioisotopes needed by the U.S. medical, industrial, and research communities.

  19. Uncertainties analysis of fission fraction for reactor antineutrino experiments

    E-Print Network [OSTI]

    X. B. Ma; F. Lu; L. Z. Wang; Y. X. Chen; W. L. Zhong; F. P. An

    2015-03-17T23:59:59.000Z

    Reactor antineutrino experiment are used to study neutrino oscillation, search for signatures of nonstandard neutrino interaction, and monitor reactor operation for safeguard application. Reactor simulation is an important source of uncertainties for a reactor neutrino experiment. Commercial code is used for reactor simulation to evaluate fission fraction in Daya Bay neutrino experiment, but the source code doesn't open to our researcher results from commercial secret. In this study, The open source code DRAGON was improved to calculate the fission rates of the four most important isotopes in fissions, $^{235}$U,$^{238}$U,$^{239}$Pu and $^{241}$Pu, and then was validated for PWRs using the Takahama-3 benchmark. The fission fraction results are consistent with those of MIT's results. Then, fission fraction of Daya Bay reactor core was calculated by using improved DRAGON code, and the fission fraction calculated by DRAGON agreed well with these calculated by SCIENCE. The average deviation less than 5\\% for all the four isotopes. The correlation coefficient matrix between $^{235}$U,$^{238}$U,$^{239}$Pu and $^{241}$Pu were also studied using DRAGON, and then the uncertainty of the antineutrino flux by the fission fraction was calculated by using the correlation coefficient matrix. The uncertainty of the antineutrino flux by the fission fraction simulation is 0.6\\% per core for Daya Bay antineutrino experiment. The uncertainties source of fission fraction calculation need further to be studied in the future.

  20. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B. (Severna Park, MD)

    2004-03-16T23:59:59.000Z

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  1. 7, 1271512750, 2007 Hydrogen isotope

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    imply that there must be a very strong concomitant isotopic enrichment in the radical channel (CH2O + hACPD 7, 12715­12750, 2007 Hydrogen isotope fractionation in the photolysis of formaldehyde T. S a Creative Commons License. Atmospheric Chemistry and Physics Discussions Hydrogen isotope fractionation

  2. Research and Medical Isotope Reactor Supply | Y-12 National Security...

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

    its predecessor. The ES 3100 accommodates many forms of highly enriched uranium and other nuclear materials in bulk quantities for both ground and air transport and uses a patented...

  3. Research and Medical Isotope Reactor Supply | Y-12 National Security

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's PossibleRadiation Protection Technical s oPrecipitationWeatherTacklingAboutNRAP:RSFComplex

  4. High Flux Isotope Reactor | Neutron Science at ORNL

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOnItem NotEnergy, science,SpeedingWu,IntelligenceYou are here ‹FIRST CenterAboutHigh Flux

  5. Atmospheric Pressure Reactor System | EMSL

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

    Atmospheric Pressure Reactor System Atmospheric Pressure Reactor System The atmospheric pressure reactor system is designed for testing the efficiency of various catalysts for the...

  6. CRDIAC: Coupled Reactor Depletion Instrument with Automated Control

    SciTech Connect (OSTI)

    Steven K. Logan

    2012-08-01T23:59:59.000Z

    When modeling the behavior of a nuclear reactor over time, it is important to understand how the isotopes in the reactor will change, or transmute, over that time. This is especially important in the reactor fuel itself. Many nuclear physics modeling codes model how particles interact in the system, but do not model this over time. Thus, another code is used in conjunction with the nuclear physics code to accomplish this. In our code, Monte Carlo N-Particle (MCNP) codes and the Multi Reactor Transmutation Analysis Utility (MRTAU) were chosen as the codes to use. In this way, MCNP would produce the reaction rates in the different isotopes present and MRTAU would use cross sections generated from these reaction rates to determine how the mass of each isotope is lost or gained. Between these two codes, the information must be altered and edited for use. For this, a Python 2.7 script was developed to aid the user in getting the information in the correct forms. This newly developed methodology was called the Coupled Reactor Depletion Instrument with Automated Controls (CRDIAC). As is the case in any newly developed methodology for modeling of physical phenomena, CRDIAC needed to be verified against similar methodology and validated against data taken from an experiment, in our case AFIP-3. AFIP-3 was a reduced enrichment plate type fuel tested in the ATR. We verified our methodology against the MCNP Coupled with ORIGEN2 (MCWO) method and validated our work against the Post Irradiation Examination (PIE) data. When compared to MCWO, the difference in concentration of U-235 throughout Cycle 144A was about 1%. When compared to the PIE data, the average bias for end of life U-235 concentration was about 2%. These results from CRDIAC therefore agree with the MCWO and PIE data, validating and verifying CRDIAC. CRDIAC provides an alternative to using ORIGEN-based methodology, which is useful because CRDIAC's depletion code, MRTAU, uses every available isotope in its depletion, unlike ORIGEN, which only depletes the isotopes specified by the user. This means that depletions done by MRTAU more accurately reflect reality. MRTAU also allows the user to build new isotope data sets, which means any isotope with nuclear data could be depleted, something that would help predict the outcomes of nuclear reaction testing in materials other than fuel, like beryllium or gold.

  7. Safe actinide disposition in molten salt reactors

    SciTech Connect (OSTI)

    Gat, U.

    1997-03-01T23:59:59.000Z

    Safe molten salt reactors (MSR) can readily accommodate the burning of all fissile actinides. Only minor compromises associated with plutonium are required. The MSRs can dispose safely of actinides and long lived isotopes to result in safer and simpler waste. Disposing of actinides in MSRs does increase the source term of a safety optimized MSR. It is concluded that the burning and transmutation of actinides in MSRs can be done in a safe manner. Development is needed for the processing to handle and separate the actinides. Calculations are needed to establish the neutron economy and the fuel management. 9 refs.

  8. Isotope Specific Remediation Media and Systems - 13614

    SciTech Connect (OSTI)

    Denton, Mark S.; Mertz, Joshua L. [Kurion, Inc. Oak Ridge, Tennessee 37831 (United States)] [Kurion, Inc. Oak Ridge, Tennessee 37831 (United States); Morita, Keisuke [Japan Atomic Energy Agency, Tokai Research and Development Center, Fukushima Project Team, Tokai-mura, Ibaraki-ken, 319-1195 (Japan)] [Japan Atomic Energy Agency, Tokai Research and Development Center, Fukushima Project Team, Tokai-mura, Ibaraki-ken, 319-1195 (Japan)

    2013-07-01T23:59:59.000Z

    On March 11, 2011, now two years ago, the magnitude 9.0 Great East Japan earthquake, Tohoku, hit off the Fukushima coast of Japan. While, of course, most of the outcome of this unprecedented natural and manmade disaster was a negative, both in Japan and worldwide, there have been some extremely invaluable lessons learned and new emergency recovery technologies and systems developed. As always, the mother of invention is necessity. Among these developments has been the development and full-scale implementation of proven isotope specific media (ISMs) with the intent of surgically removing specific hazardous isotopes for the purpose of minimizing dose to workers and the environment. The first such ISMs to be deployed at the Fukushima site were those removing cesium (Cs-137) and iodine (I-129). Since deployment on June 17, 2011, along with treated cooling water recycle, some 70% of the curies in the building liquid wastes have been removed by the Kurion system alone. The current levels of cesium are now only 2% of the original levels. Such an unprecedented, 'external cooling system' not only allowed the eventual cold shut down of the reactors in mid-December, 2011, but has allowed workers to concentrate on the cleanup of other areas of the site. Water treatment will continue for quite some time due to continued leakage into the buildings and the eventual goal of cleaning up the reactors and fuel pools themselves. With the cesium removal now in routine operation, other isotopes of concern are likely to become priorities. One such isotope is that of strontium, and yttrium (Sr-90 and Y-90), which is still at original levels causing further dose issues as well as impediments to discharge of the treated waste waters. For over a year now, a new synthetic strontium specific media has been under development and testing both in our licensed facility in Oak Ridge, Tennessee, but also in confirmatory tests by the Japan Atomic Energy Agency (JAEA) in Japan for Tokyo Electric Power Company (TEPCO). The tests have proven quite successful, even in high salt conditions, and, with loading and dose calculations being completed, will be proposed to add to the existing cesium system. There is no doubt, as high gamma isotopes are removed, other recalcitrant isotopes such as this will require innovative removal media, systems and techniques. Also coming out of this international effort are other ISM media and systems that can be applied more broadly to both Commercial Nuclear Power Plants (NPPs) as well as in Department of Energy (DOE) applications. This cesium and strontium specific media has further been successfully tested in 2012 at a Magnox station in the UK. The resulting proposed mitigation systems for pond and vault cleanup look quite promising. An extremely unusual ISM for carbon 14 (C-14), nickel (Ni-63) and cesium (Cs-137) has been developed for Diablo Canyon NPP for dose reduction testing in their fuel pool. These media will be deployed in Submersible Media Filter (SMF) and Submersible Columns (SC) systems adapted to standard Tri-Nuclear{sup R} housings common in the U.S. and UK. External Vessel Systems (mini-Fukushima) have also been developed as a second mitigation system for D and D and outages. Finally, technetium (Tc- 99) specific media developed for the Waste Treatment Plant (WTP) recycle or condensate (secondary) waste streams (WM 2011) are being further perfected and tested for At-Tank Tc-99 removal, as well as At Tank Cs media. In addition to the on-going media development, systems for deploying such media have developed over the last year and are in laboratory- and full-scale testing. These systems include the fore mentioned Submersible Media Filters (SMF), Submersible Columns (SC) and external pilot- and full-scale, lead-lag, canister systems. This paper will include the media development and testing, as well as that of the deployment systems themselves. (authors)

  9. Reactor Sharing Program

    SciTech Connect (OSTI)

    Vernetson, W.G.

    1993-01-01T23:59:59.000Z

    Progress achieved at the University of Florida Training Reactor (UFTR) facility through the US Department of Energy's University Reactor Sharing Program is reported for the period of 1991--1992.

  10. Undergraduate reactor control experiment

    SciTech Connect (OSTI)

    Edwards, R.M.; Power, M.A.; Bryan, M. (Pennsylvania State Univ., University Park (United States))

    1992-01-01T23:59:59.000Z

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise.

  11. Attrition reactor system

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)

    1993-01-01T23:59:59.000Z

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  12. Attrition reactor system

    DOE Patents [OSTI]

    Scott, C.D.; Davison, B.H.

    1993-09-28T23:59:59.000Z

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  13. ISOTOPE METHODS IN HOMOGENEOUS CATALYSIS.

    SciTech Connect (OSTI)

    BULLOCK,R.M.; BENDER,B.R.

    2000-12-01T23:59:59.000Z

    The use of isotope labels has had a fundamentally important role in the determination of mechanisms of homogeneously catalyzed reactions. Mechanistic data is valuable since it can assist in the design and rational improvement of homogeneous catalysts. There are several ways to use isotopes in mechanistic chemistry. Isotopes can be introduced into controlled experiments and followed where they go or don't go; in this way, Libby, Calvin, Taube and others used isotopes to elucidate mechanistic pathways for very different, yet important chemistries. Another important isotope method is the study of kinetic isotope effects (KIEs) and equilibrium isotope effect (EIEs). Here the mere observation of where a label winds up is no longer enough - what matters is how much slower (or faster) a labeled molecule reacts than the unlabeled material. The most careti studies essentially involve the measurement of isotope fractionation between a reference ground state and the transition state. Thus kinetic isotope effects provide unique data unavailable from other methods, since information about the transition state of a reaction is obtained. Because getting an experimental glimpse of transition states is really tantamount to understanding catalysis, kinetic isotope effects are very powerful.

  14. Laser isotope separation of erbium and other isotopes

    DOE Patents [OSTI]

    Haynam, Christopher A. (3035 Ferdale Ct., Pleasanton, CA 94566); Worden, Earl F. (117 Vereda del Ciervo, Diablo, CA 94528)

    1995-01-01T23:59:59.000Z

    Laser isotope separation is accomplished using at least two photoionization pathways of an isotope simultaneously, where each pathway comprises two or more transition steps. This separation method has been applied to the selective photoionization of erbium isotopes, particularly for the enrichment of .sup.167 Er. The hyperfine structure of .sup.167 Er was used to find two three-step photoionization pathways having a common upper energy level.

  15. Laser isotope separation of erbium and other isotopes

    DOE Patents [OSTI]

    Haynam, C.A.; Worden, E.F.

    1995-08-22T23:59:59.000Z

    Laser isotope separation is accomplished using at least two photoionization pathways of an isotope simultaneously, where each pathway comprises two or more transition steps. This separation method has been applied to the selective photoionization of erbium isotopes, particularly for the enrichment of {sup 167}Er. The hyperfine structure of {sup 167}Er was used to find two three-step photoionization pathways having a common upper energy level. 3 figs.

  16. Advanced Test Reactor Tour

    SciTech Connect (OSTI)

    Miley, Don

    2011-01-01T23:59:59.000Z

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  17. High solids fermentation reactor

    DOE Patents [OSTI]

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-03-02T23:59:59.000Z

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  18. High solids fermentation reactor

    DOE Patents [OSTI]

    Wyman, Charles E. (Lakewood, CO); Grohmann, Karel (Littleton, CO); Himmel, Michael E. (Littleton, CO); Richard, Christopher J. (Lakewood, CO)

    1993-01-01T23:59:59.000Z

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  19. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

    1995-01-01T23:59:59.000Z

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  20. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28T23:59:59.000Z

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  1. SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

    SciTech Connect (OSTI)

    J.W. Davis

    1996-08-29T23:59:59.000Z

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

  2. Tritium Formation and Mitigation in High-Temperature Reactor Systems

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

    2013-03-01T23:59:59.000Z

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  3. Tritium Formation and Mitigation in High-Temperature Reactors

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots

    2012-10-01T23:59:59.000Z

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  4. Water isotopes and the general circulation

    E-Print Network [OSTI]

    Noone, David

    is depleted. #12;Distillation: vapor and condensate Isotopic fractionation -35 -30 -25 -20 -15 -10 -5 0 5 0 0 of idealized isotopic fractionation Expression of isotopic fractionation in nature Attributing signals" 18 = (R/Rstandard-1)x1000 R = moles of H2 18O/moles of H2 16O #12;Isotopic fractionation Isotopic

  5. Neutron behavior, reactor control, and reactor heat transfer. Volume four

    SciTech Connect (OSTI)

    Not Available

    1986-01-01T23:59:59.000Z

    Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant).

  6. Reactor vessel support system

    DOE Patents [OSTI]

    Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

    1982-01-01T23:59:59.000Z

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  7. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, D.M.; Taft, W.E.

    1994-12-20T23:59:59.000Z

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  8. Spinning fluids reactor

    SciTech Connect (OSTI)

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20T23:59:59.000Z

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  9. Phoenix: A Reactor Burnup Code With Uncertainty Quantification 

    E-Print Network [OSTI]

    Spence, Grant R

    2014-12-15T23:59:59.000Z

    composition, and cross-section, while only running the simulation a single time. This ability provides an enormous benefit to researchers using the code for spent fuel reprocessing, safeguards, nuclear nonproliferation, and intelligence applications. 6... reprocessing, reactor operation, and many nuclear intelligence applications. Developing burnup software, like any computational software, 2 can be a difficult process. It requires defining the parameters necessary to solve isotopic depletion equations...

  10. Novel Isotope Effects and Organic Reaction Mechanisms

    E-Print Network [OSTI]

    Kelly, Kelmara K.

    2010-07-14T23:59:59.000Z

    to account for the observed isotope effects. In the dimerization of cyclopentadiene, novel "dynamic" isotope effects are observed on the 13C distribution in the product, and a method for the prediction of these isotope effects is developed here...

  11. Gas tagging and cover gas combination for nuclear reactor

    DOE Patents [OSTI]

    Gross, Kenny C. (Lemont, IL); Laug, Matthew T. (Idaho Falls, ID)

    1985-01-01T23:59:59.000Z

    The invention discloses the use of stable isotopes of neon and argon, that are grouped in preselected different ratios one to the other and are then sealed as tags in different cladded nuclear fuel elements to be used in a liquid metal fast breeder reactor. Failure of the cladding of any fuel element allows fission gases generated in the reaction and these tag isotopes to escape and to combine with the cover gas held in the reactor over the fuel elements. The isotopes specifically are Ne.sup.20, Ne.sup.21 and Ne.sup.22 of neon and Ar.sup.36, Ar.sup.38 and Ar.sup.40 of argon, and the cover gas is helium. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between approximately 0.degree. and -25.degree. C. operable to remove the fission gases from the cover gas and tags and the second or tag recovery system bed is held between approximately -170.degree. and -185.degree. C. operable to isolate the tags from the cover gas. Spectrometric analysis further is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be specifically determined.

  12. ARM - Measurement - Isotope ratio

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625govInstrumentstdmadap Documentation TDMADAP : XDC documentationBarrow,ice particleSizegovMeasurementsIsotope ratio

  13. The marine biogeochemistry of zinc isotopes

    E-Print Network [OSTI]

    John, Seth G

    2007-01-01T23:59:59.000Z

    Zinc (Zn) stable isotopes can record information about important oceanographic processes. This thesis presents data on Zn isotopes in anthropogenic materials, hydrothermal fluids and minerals, cultured marine phytoplankton, ...

  14. Stable Isotope Fractionations in Biogeochemical Reactive Transport

    E-Print Network [OSTI]

    Druhan, Jennifer Lea

    2012-01-01T23:59:59.000Z

    characteristic of stable isotope enrichment. The values of !isotope ratios of sulfur in these sulfate samples demonstrated a clear enrichmentisotope ( 34 S) (Canfield, 2001). The characteristic enrichment

  15. Strategic Isotope Production | ornl.gov

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

    Strategic Isotope Production SHARE Strategic Isotope Production Typical capsules used in the transport of 252Cf source material inside heavily shielded shipping casks. ORNL's...

  16. Isotopic Trends in Production of Superheavies

    SciTech Connect (OSTI)

    Antonenko, N.V. [Institut fuer Theoretische Physik der Justus-Liebig-Universitaet, D-35392 Giessen (Germany); Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Adamian, G.G.; Zubov, A.S. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Scheid, W. [Institut fuer Theoretische Physik der Justus-Liebig-Universitaet, D-35392 Giessen (Germany)

    2005-11-21T23:59:59.000Z

    The isotopic trends are discussed for cold and hot fusion reactions leading to superheavies. The possibilities of production of new isotopes in incomplete fusion reactions are treated.

  17. Isotope Research 229 Th production

    E-Print Network [OSTI]

    Isotope Research ­ 229 Th production We recently completed an ARRA-funded project of this type on 229 Th production reactions [Str11]. This long-lived isotope is important as a precursor to 225 Ac of accelerator production of 229 Th via the 230 Th(p,2n)229 Pa reaction. The 229 Pa decays primarily by electron

  18. Method of separating boron isotopes

    DOE Patents [OSTI]

    Jensen, Reed J. (Los Alamos, NM); Thorne, James M. (Provo, UT); Cluff, Coran L. (Provo, UT); Hayes, John K. (Salt Lake City, UT)

    1984-01-01T23:59:59.000Z

    A method of boron isotope enrichment involving the isotope preferential photolysis of (2-chloroethenyl)dichloroborane as the feed material. The photolysis can readily be achieved with CO.sub.2 laser radiation and using fluences significantly below those required to dissociate BCl.sub.3.

  19. Method of separating boron isotopes

    DOE Patents [OSTI]

    Jensen, R.J.; Thorne, J.M.; Cluff, C.L.

    1981-01-23T23:59:59.000Z

    A method of boron isotope enrichment involving the isotope preferential photolysis of (2-chloroethenyl)-dichloroborane as the feed material. The photolysis can readily by achieved with CO/sub 2/ laser radiation and using fluences significantly below those required to dissociate BCl/sub 3/.

  20. Implications of Plutonium isotopic separation on closed fuel cycles and repository design

    SciTech Connect (OSTI)

    Forsberg, C. [Massachusetts Institute of Technology, 77 Massachusetts Ave. Cambridge, MA 20129 (United States)

    2013-07-01T23:59:59.000Z

    Advances in laser enrichment may enable relatively low-cost plutonium isotopic separation. This would have large impacts on LWR closed fuel cycles and waste management. If Pu-240 is removed before recycling plutonium as mixed oxide (MOX) fuel, it would dramatically reduce the buildup of higher plutonium isotopes, Americium, and Curium. Pu-240 is a fertile material and thus can be replaced by U-238. Eliminating the higher plutonium isotopes in MOX fuel increases the Doppler feedback, simplifies reactor control, and allows infinite recycle of MOX plutonium in LWRs. Eliminating fertile Pu-240 and Pu-242 reduces the plutonium content in MOX fuel and simplifies fabrication. Reducing production of Pu-241 reduces production of Am-241 - the primary heat generator in spent nuclear fuels after several decades. Reducing heat generating Am-241 would reduce repository cost and waste toxicity. Avoiding Am- 241 avoids its decay product Np-237, a nuclide that partly controls long-term oxidizing repository performance. Most of these benefits also apply to LWR plutonium recycled into fast reactors. There are benefits for plutonium isotopic separation in fast reactor fuel cycles (particularly removal of Pu-242) but the benefits are less. (author)

  1. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Timothy A. Hyde

    2012-06-01T23:59:59.000Z

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  2. Isotope separation by laser means

    DOE Patents [OSTI]

    Robinson, C. Paul (Los Alamos, NM); Jensen, Reed J. (Los Alamos, NM); Cotter, Theodore P. (Los Alamos, NM); Greiner, Norman R. (Los Alamos, NM); Boyer, Keith (Los Alamos, NM)

    1982-06-15T23:59:59.000Z

    A process for separating isotopes by selective excitation of isotopic species of a volatile compound by tuned laser light. A highly cooled gas of the volatile compound is produced in which the isotopic shift is sharpened and defined. Before substantial condensation occurs, the cooled gas is irradiated with laser light precisely tuned to a desired wavelength to selectively excite a particular isotopic species in the cooled gas. The laser light may impart sufficient energy to the excited species to cause it to undergo photochemical reaction or even to photoionize. Alternatively, a two-photon irradiation may be applied to the cooled gas to induce photochemical reaction or photoionization. The process is particularly applicable to the separation of isotopes of uranium and plutonium.

  3. Gamma ray-induced embrittlement of pressure vessel alloys

    SciTech Connect (OSTI)

    Alexander, D.E.; Rehn, L.E. [Argonne National Lab., IL (United States); Farrell, K.; Stoller, R.E. [Oak Ridge National Lab., TN (United States)

    1994-11-01T23:59:59.000Z

    High-energy gamma rays emitted from the core of a nuclear reactor produce displacement damage in the reactor pressure vessel (RPV). The contribution of gamma damage to RPV embrittlement has in the past been largely ignored. However, in certain reactor designs the gamma flux at the RPV is sufficiently large that its contribution to displacement damage can be substantial. For example, gamma rays have been implicated in the accelerated RPV embrittlement observed in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. In the present study, mechanical property changes induced by 10-MeV electron irradiation of a model Fe alloy and an RPV alloy of interest to the HFIR were examined. Mini-tensile specimens were irradiated with high-energy electrons to reproduce damage characteristic of the Compton recoil-electrons induced by gamma bombardment. Substantial increases in yield and ultimate stress were observed in the alloys after irradiation to doses up to 5.3x10{sup {minus}3} dpa at temperatures ({approximately}50{degrees}C) characteristic of the HFIR pressure vessel. These measured increases were similar to those previously obtained following neutron irradiation, despite the highly disparate nature of the damage generated during electron and neutron irradiation.

  4. Reactor Physics Assessment of the Inclusion of Unseparated Neptunium in MOX Reactor Fuel

    SciTech Connect (OSTI)

    Ellis, Ronald James [ORNL

    2009-01-01T23:59:59.000Z

    Reducing the number of actinide separation streams in a spent fuel recovery process would reduce the cost and complexity of the process, and lower the quantity and numbers of solvents needed. It is more difficult and costly to separate Np and recombine it with Am-Cm prior to co-conversion than to simply co-strip it with the U-Pu-Np. Inclusion of the Np in mixed oxide (MOX) fuel for light water reactor (LWR) applications should not seriously affect the operating behavior of the reactor, nor should it pose insurmountable fuel design issues. In this work, the U, Pu, and Np from typical discharged and cooled PWR spent nuclear fuel are assumed to be used together in the preparation of MOX fuel for use in a pressurized water reactor (PWR). The reactor grade Pu isotopic vector is used in the model and the relative mass ratio of the Pu and Np content (Np/Pu mass is 0.061) from the cooled spent fuel is maintained but the overall Pu-Np MOX wt% is adjusted with respect to the U content (assumed to be at 0.25 wt% 235U enrichment) to offset reactivity and cycle length effects. The SCALE 5.1 scientific package (especially modules TRITON, NEWT, ORIGEN-S, ORIGEN-ARP) was used for the calculations presented in this paper. A typical Westinghouse 17x17 fuel assembly design was modeled at nominal PWR operating conditions. It was seen that U-Pu-Np MOX fuel with NpO2 and PuO2 representing 11.5wt% of the total MOX fuel would be similar to standard MOX fuel in which PuO2 is 9wt% of the fuel. The reactivity, isotopic composition, and neutron and ? sources, and the decay heat details for the discharged MOX fuel are presented and discussed in this paper.

  5. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, J.P.; Scahill, J.W.

    1995-05-09T23:59:59.000Z

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  6. Tokamak reactor first wall

    DOE Patents [OSTI]

    Creedon, R.L.; Levine, H.E.; Wong, C.; Battaglia, J.

    1984-11-20T23:59:59.000Z

    This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

  7. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, Juhani (Karhula, FI)

    1996-01-01T23:59:59.000Z

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

  8. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, J.

    1996-03-19T23:59:59.000Z

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

  9. Analysis of a research reactor under anticipated transients without scram events using the RELAP5/MOD3.2 computer program

    E-Print Network [OSTI]

    Hari, Sridhar

    1998-01-01T23:59:59.000Z

    . . . 54 III Primary Loop Parameters: Comparison of Research Reactor With a Typical PWR. IV Summary of the Results of the Simulated Transients. . . 57 93 ACRONYMS AAEC ANL Australian Atomic Energy Commission Argonne National Laboratory ANSTO... Basis Accident Emergency Core Cooling System High Flux Australian Reactor HIFAR specific version of the ZAPP code High Flux Isotope Reactor International Atomic Energy Agency Idaho National Engineering Laboratory Loss of Coolant Accident Critical...

  10. NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE

    SciTech Connect (OSTI)

    John D. Bess

    2014-03-01T23:59:59.000Z

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately +/-1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  11. Oklo reactors and implications for nuclear science

    E-Print Network [OSTI]

    E. D. Davis; C. R. Gould; E. I. Sharapov

    2014-04-19T23:59:59.000Z

    We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlier assumed. Nuclear cross sections are input to all Oklo modeling and we discuss a parameter, the $^{175}$Lu ground state cross section for thermal neutron capture leading to the isomer $^{176\\mathrm{m}}$ Lu, that warrants further investigation. Studies of the time dependence of dimensionless fundamental constants have been a driver for much of the recent work on Oklo. We critically review neutron resonance energy shifts and their dependence on the fine structure constant $\\alpha$ and the ratio $X_q=m_q/\\Lambda$ (where $m_q$ is the average of the $u$ and $d$ current quark masses and $\\Lambda$ is the mass scale of quantum chromodynamics). We suggest a formula for the combined sensitivity to $\\alpha$ and $X_q$ that exhibits the dependence on proton number $Z$ and mass number $A$, potentially allowing quantum electrodynamic and quantum chromodynamic effects to be disentangled if a broader range of isotopic abundance data becomes available.

  12. Developing the Sandia National Laboratories transportation infrastructure for isotope products and wastes

    SciTech Connect (OSTI)

    Trennel, A.J.

    1997-11-01T23:59:59.000Z

    The US Department of Energy (DOE) plans to establish a medical isotope project that would ensure a reliable domestic supply of molybdenum-99 ({sup 99}Mo) and related medical isotopes (Iodine-125, Iodine-131, and Xenon-133). The Department`s plan for production will modify the Annular Core Research Reactor (ACRR) and associated hot cell facility at Sandia National Laboratories (SNL)/New Mexico and the Chemistry and Metallurgy Research facility at Los Alamos National Laboratory (LANL). Transportation activities associated with such production is discussed.

  13. Brookhaven Graphite Research Reactor Workshop

    Broader source: Energy.gov [DOE]

    The Brookhaven Graphite Research Reactor (BGRR) was the first reactor built in the U.S. for peacetime atomic research following World War II.  Construction began in 1947 and the reactor started...

  14. Apparatus and process for separating hydrogen isotopes

    DOE Patents [OSTI]

    Heung, Leung K; Sessions, Henry T; Xiao, Xin

    2013-06-25T23:59:59.000Z

    The apparatus and process for separating hydrogen isotopes is provided using dual columns, each column having an opposite hydrogen isotopic effect such that when a hydrogen isotope mixture feedstock is cycled between the two respective columns, two different hydrogen isotopes are separated from the feedstock.

  15. Method for laser induced isotope enrichment

    DOE Patents [OSTI]

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

    2004-09-07T23:59:59.000Z

    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.

  16. INITIAL IRRADIATION OF THE FIRST ADVANCED GAS REACTOR FUEL DEVELOPMENT AND QUALIFICATION EXPERIMENT IN THE ADVANCED TEST REACTOR

    SciTech Connect (OSTI)

    S. Blaine Grover; David A. Petti

    2007-09-01T23:59:59.000Z

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  17. Compelling Research Opportunities using Isotopes

    SciTech Connect (OSTI)

    None

    2009-04-23T23:59:59.000Z

    Isotopes are vital to the science and technology base of the US economy. Isotopes, both stable and radioactive, are essential tools in the growing science, technology, engineering, and health enterprises of the 21st century. The scientific discoveries and associated advances made as a result of the availability of isotopes today span widely from medicine to biology, physics, chemistry, and a broad range of applications in environmental and material sciences. Isotope issues have become crucial aspects of homeland security. Isotopes are utilized in new resource development, in energy from bio-fuels, petrochemical and nuclear fuels, in drug discovery, health care therapies and diagnostics, in nutrition, in agriculture, and in many other areas. The development and production of isotope products unavailable or difficult to get commercially have been most recently the responsibility of the Department of Energy's Nuclear Energy program. The President's FY09 Budget request proposed the transfer of the Isotope Production program to the Department of Energy's Office of Science in Nuclear Physics and to rename it the National Isotope Production and Application program (NIPA). The transfer has now taken place with the signing of the 2009 appropriations bill. In preparation for this, the Nuclear Science Advisory Committee (NSAC) was requested to establish a standing subcommittee, the NSAC Isotope Subcommittee (NSACI), to advise the DOE Office of Nuclear Physics. The request came in the form of two charges: one, on setting research priorities in the short term for the most compelling opportunities from the vast array of disciplines that develop and use isotopes and two, on making a long term strategic plan for the NIPA program. This is the final report to address charge 1. NSACI membership is comprised of experts from the diverse research communities, industry, production, and homeland security. NSACI discussed research opportunities divided into three areas: (1) medicine, pharmaceuticals, and biology, (2) physical sciences and engineering, and (3) national security and other applications. In each area, compelling research opportunities were considered and the subcommittee as a whole determined the final priorities for research opportunities as the foundations for the recommendations. While it was challenging to prioritize across disciplines, our order of recommendations reflect the compelling research prioritization along with consideration of time urgency for action as well as various geopolitical market issues. Common observations to all areas of research include the needs for domestic availability of crucial stable and radioactive isotopes and the education of the skilled workforce that will develop new advances using isotopes in the future. The six recommendations of NSACI reflect these concerns and the compelling research opportunities for potential new discoveries. The science case for each of the recommendations is elaborated in the respective chapters.

  18. Electrochemical Isotope Effect and Lithium Isotope Separation Jay R. Black,

    E-Print Network [OSTI]

    Mcdonough, William F.

    results showing a large lithium isotope separation due to electrodeposition. The fractionation is tunable lithium were plated from solutions of 1 M LiClO4 in propylene carbonate (PC) on planar nickel electrodes

  19. Isotopic Tracer Studies of Propane Reactions on H-ZSM5 Zeolite Joseph A. Biscardi and Enrique Iglesia*

    E-Print Network [OSTI]

    Iglesia, Enrique

    Isotopic Tracer Studies of Propane Reactions on H-ZSM5 Zeolite Joseph A. Biscardi and Enrique unlabeled products from mixtures of propene and propane-2-13C reactants. Aromatic products of propane-2-13C-Parmer) that allowed differential reactor operation (propane reactions were

  20. REACTOR OPERATIONS AND CONTROL

    E-Print Network [OSTI]

    Pázsit, Imre

    REACTOR OPERATIONS AND CONTROL KEYWORDS: core calculations, neural networks, control rod elevation of a control rod, or a group of control rods, is an important parameter from the viewpoint of reactor control DETERMINATION OF PWR CONTROL ROD POSITION BY CORE PHYSICS AND NEURAL NETWORK METHODS NINOS S. GARIS* and IMRE

  1. Reed Reactor Facility. Final report

    SciTech Connect (OSTI)

    Frantz, S.G.

    1994-12-31T23:59:59.000Z

    This report discusses the operation and maintenance of the Reed Reactor Facility. The Reed reactor is mostly used for education and train purposes.

  2. Reactor & Nuclear Systems Publications | ORNL

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

    Science Home | Science & Discovery | Nuclear Science | Publications and Reports | Reactor and Nuclear Systems Publications SHARE Reactor and Nuclear Systems Publications The...

  3. Novel hybrid isotope separation scheme and apparatus

    DOE Patents [OSTI]

    Maya, Jakob (Brookline, MA)

    1991-01-01T23:59:59.000Z

    A method of yielding selectively a desired enrichment in a specific isotope including the steps of inputting into a spinning chamber a gas from which the specific isotope is to be isolated, radiating the gas with frequencies characteristic of the absorption of a particular isotope of the atomic or molecular gas, thereby inducing a photoionization reaction of the desired isotope, and collecting the specific isotope ion by suitable ion collection means.

  4. Novel hybrid isotope separation scheme and apparatus

    DOE Patents [OSTI]

    Maya, J.

    1991-06-18T23:59:59.000Z

    A method is described for yielding selectively a desired enrichment in a specific isotope including the steps of inputting into a spinning chamber a gas from which the specific isotope is to be isolated, radiating the gas with frequencies characteristic of the absorption of a particular isotope of the atomic or molecular gas, thereby inducing a photoionization reaction of the desired isotope, and collecting the specific isotope ion by suitable ion collection means. 3 figures.

  5. The Origin and Implications of the Shoulder in Reactor Neutrino Spectra

    E-Print Network [OSTI]

    A. C. Hayes; J. L. Friar; G. T. Garvey; Duligur Ibeling; Gerard Jungman; T. Kawano; Robert W. Mills

    2015-06-01T23:59:59.000Z

    We analyze within a nuclear database framework the shoulder observed in the antineutrino spectra in current reactor experiments. We find that the ENDF/B-VII.1 database predicts that the antineutrino shoulder arises from an analogous shoulder in the aggregate fission beta spectra. In contrast, the JEFF-3.1.1 database does not predict a shoulder. We consider several possible origins of the shoulder, and find possible explanations. For example, there could be a problem with the measured aggregate beta spectra, or the harder neutron spectrum at a light-water power reactor could affect the distribution of beta-decaying isotopes. In addition to the fissile actinides, we find that $^{238}$U could also play a significant role in distorting the total antineutrino spectrum. Distinguishing these and quantifying whether there is an anomaly associated with measured reactor neutrino signals will require new short-baseline experiments, both at thermal reactors and at reactors with a sizable epithermal neutron component.

  6. The Origin and Implications of the Shoulder in Reactor Neutrino Spectra

    E-Print Network [OSTI]

    Hayes, A C; Garvey, G T; Ibeling, Duligur; Jungman, Gerard; Kawano, T; Mills, Robert W

    2015-01-01T23:59:59.000Z

    We analyze within a nuclear database framework the shoulder observed in the antineutrino spectra in current reactor experiments. We find that the ENDF/B-VII.1 database predicts that the antineutrino shoulder arises from an analogous shoulder in the aggregate fission beta spectra. In contrast, the JEFF-3.1.1 database does not predict a shoulder. We consider several possible origins of the shoulder, and find possible explanations. For example, there could be a problem with the measured aggregate beta spectra, or the harder neutron spectrum at a light-water power reactor could affect the distribution of beta-decaying isotopes. In addition to the fissile actinides, we find that $^{238}$U could also play a significant role in distorting the total antineutrino spectrum. Distinguishing these and quantifying whether there is an anomaly associated with measured reactor neutrino signals will require new short-baseline experiments, both at thermal reactors and at reactors with a sizable epithermal neutron component.

  7. Interim Safe Storage of Plutonium Production Reactors at the US DOE Hanford Site - 13438

    SciTech Connect (OSTI)

    Schilperoort, Daryl L.; Faulk, Darrin [Washington Closure Hanford, 2620 Fermi Avenue, Richland, Washington 99352 (United States)] [Washington Closure Hanford, 2620 Fermi Avenue, Richland, Washington 99352 (United States)

    2013-07-01T23:59:59.000Z

    Nine plutonium production reactors located on DOE's Hanford Site are being placed into an Interim Safe Storage (ISS) period that extends to 2068. The Environmental Impact Statement (EIS) for ISS [1] was completed in 1993 and proposed a 75-year storage period that began when the EIS was finalized. Remote electronic monitoring of the temperature and water level alarms inside the safe storage enclosure (SSE) with visual inspection inside the SSE every 5 years are the only planned operational activities during this ISS period. At the end of the ISS period, the reactor cores will be removed intact and buried in a landfill on the Hanford Site. The ISS period allows for radioactive decay of isotopes, primarily Co-60 and Cs-137, to reduce the dose exposure during disposal of the reactor cores. Six of the nine reactors have been placed into ISS by having an SSE constructed around the reactor core. (authors)

  8. Selective photoionisation of lutetium isotopes

    SciTech Connect (OSTI)

    D'yachkov, Aleksei B; Kovalevich, S K; Labozin, Valerii P; Mironov, Sergei M; Panchenko, Vladislav Ya; Firsov, Valerii A; Tsvetkov, G O; Shatalova, G G [National Research Centre 'Kurchatov Institute', Moscow (Russian Federation)

    2012-10-31T23:59:59.000Z

    A three-stage laser photoionisation scheme intended for enriching the {sup 176}Lu isotope from natural lutetium was considered. An investigation was made of the hyperfine structure of the second excited state 5d6s7s {yields} {sup 4}D{sub 3/2} with an energy of 37194 cm{sup -1} and the autoionisation state with an energy of 53375 cm{sup -1} of the {sup 176}Lu and {sup 175}Lu isotopes. The total electron momentum of the autoionisation level and the constant A of hyperfine magnetic interaction were determined. Due to a small value of the isotopic shift between {sup 176}Lu and {sup 175}Lu, appreciable selectivity of their separation may be achieved with individual hyperfine structure components. The first tentative enrichment of the 176Lu isotope was performed to a concentration of 60 % - 70 %. (laser applications and other topics in quantum electronics)

  9. Physics with isotopically controlled semiconductors

    SciTech Connect (OSTI)

    Haller, E. E., E-mail: eehaller@lbl.gov [University of California at Berkeley, Department of Materials Science and Engineering (United States)

    2010-07-15T23:59:59.000Z

    This paper is based on a tutorial presentation at the International Conference on Defects in Semiconductors (ICDS-25) held in Saint Petersburg, Russia in July 2009. The tutorial focused on a review of recent research involving isotopically controlled semiconductors. Studies with isotopically enriched semiconductor structures experienced a dramatic expansion at the end of the Cold War when significant quantities of enriched isotopes of elements forming semiconductors became available for worldwide collaborations. Isotopes of an element differ in nuclear mass, may have different nuclear spins and undergo different nuclear reactions. Among the latter, the capture of thermal neutrons which can lead to neutron transmutation doping, is the most prominent effect for semiconductors. Experimental and theoretical research exploiting the differences in all the properties has been conducted and will be illustrated with selected examples.

  10. Retention of Hydrogen Isotopes in Neutron Irradiated Tungsten

    SciTech Connect (OSTI)

    Yuji Hatano; Masashi Shimada; Yasuhisa Oya; Guoping Cao; Makoto Kobayashi; Masanori Hara; Brad J. Merrill; Kenji Okuno; Mikhail A. Sokolov; Yutai Katoh

    2013-03-01T23:59:59.000Z

    To investigate the effects of neutron irradiation on hydrogen isotope retention in tungsten, disk-type specimens of pure tungsten were irradiated in the High Flux Isotope Reactor in Oak Ridge National Laboratory followed by exposure to high flux deuterium (D) plasma in Idaho National Laboratory. The results obtained for low dose n-irradiated specimens (0.025 dpa for tungsten) are reviewed in this paper. Irradiation at coolant temperature of the reactor (around 50 degrees C) resulted in the formation of strong trapping sites for D atoms. The concentrations of D in n-irradiated specimens were ranging from 0.1 to 0.4 mol% after exposure to D plasma at 200 and 500 degrees C and significantly higher than those in non-irradiated specimens because of D-trapping by radiation defects. Deep penetration of D up to a depth of 50-100 µm was observed at 500 degrees C. Release of D in subsequent thermal desorption measurements continued up to 900 degrees C. These results were compared with the behaviour of D in ion-irradiated tungsten, and distinctive features of n-irradiation were discussed.

  11. Isotope production facility produces cancer-fighting actinium

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

    Cancer therapy gets a boost from new isotope Isotope production facility produces cancer-fighting actinium A new medical isotope project shows promise for rapidly producing major...

  12. Nuclear reactor control column

    SciTech Connect (OSTI)

    Bachovchin, D.M.

    1982-08-10T23:59:59.000Z

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  13. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Edler, S. K.

    1981-07-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  14. Nuclear reactor control column

    DOE Patents [OSTI]

    Bachovchin, Dennis M. (Plum Borough, PA)

    1982-01-01T23:59:59.000Z

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  15. Process for preparing a chemical compound enriched in isotope content

    DOE Patents [OSTI]

    Michaels, Edward D. (Spring Valley, OH)

    1982-01-01T23:59:59.000Z

    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.

  16. Fast Breeder Reactor studies

    SciTech Connect (OSTI)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01T23:59:59.000Z

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  17. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Martin Peng, Y.K.M.

    1985-10-03T23:59:59.000Z

    The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

  18. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

    1994-01-01T23:59:59.000Z

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

  19. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, R.J.; Land, J.T.; Misvel, M.C.

    1994-06-07T23:59:59.000Z

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

  20. Microfluidic electrochemical reactors

    DOE Patents [OSTI]

    Nuzzo, Ralph G. (Champaign, IL); Mitrovski, Svetlana M. (Urbana, IL)

    2011-03-22T23:59:59.000Z

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  1. Improved gas tagging and cover gas combination for nuclear reactor

    DOE Patents [OSTI]

    Gross, K.C.; Laug, M.T.

    1983-09-26T23:59:59.000Z

    The invention discloses the use of stable isotopes of neon and argon, sealed as tags in different cladding nuclear fuel elements to be used in a liquid metal fast breeder reactor. Cladding failure allows fission gases and these tag isotopes to escape and to combine with the cover gas. The isotopes are Ne/sup 20/, Ne/sup 21/ and Ne/sup 22/ and Ar/sup 36/, Ar/sup 38/ and Ar/sup 40/, and the cover gas is He. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between 0 and -25/sup 0/C to remove the fission gases from the cover gas and tags, and the second or tag recovery system bed between -170 and -185/sup 0/C to isolate the tags from the cover gas. Spectrometric analysis is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be determined.

  2. An instrumental and numerical method to determine the hydrogenic ratio in isotopic experiments in the TJ-II stellarator

    SciTech Connect (OSTI)

    Baciero, A., E-mail: alfonso.baciero@ciemat.es; Zurro, B. [Laboratorio Nacional de Fusión, CIEMAT, Madrid (Spain); Martínez, M. [Departamento de Física, Universidad Carlos III de Madrid, Leganés (Spain)

    2014-11-15T23:59:59.000Z

    The isotope effect is an important topic that is relevant for future D-T fusion reactors, where the use of deuterium, rather than hydrogen, may lean to improved plasma confinement. An evaluation of the ratio of hydrogen/deuterium is needed for isotope effect studies in current isotopic experiments. Here, the spectral range around H{sub ?} and D{sub ?} lines, obtained with an intensified multi-channel detector mounted to a 1-m focal length spectrometer, is analyzed using a fit function that includes several Gaussian components. The isotopic ratio evolution for a single operational day of the TJ-II stellarator is presented. The role of injected hydrogen by Neutral Beam Injection heating is also studied.

  3. Improved predictions of reactor antineutrino spectra

    SciTech Connect (OSTI)

    Mueller, Th. A.; Lhuillier, D.; Letourneau, A. [Commissariat a l'Energie Atomique et aux Energies Alternatives, Centre de Saclay, IRFU/SPhN, FR-91191 Gif-sur-Yvette (France); Fallot, M.; Cormon, S.; Giot, L.; Martino, J.; Porta, A.; Yermia, F. [Laboratoire SUBATECH, Ecole des Mines de Nantes, Universite de Nantes, CNRS/IN2P3, 4 rue Alfred Kastler, FR-44307 Nantes Cedex 3 (France); Fechner, M.; Lasserre, T.; Mention, G. [Commissariat a l'Energie Atomique et aux Energies Alternatives, Centre de Saclay, IRFU/SPP, FR-91191 Gif-sur-Yvette (France)

    2011-05-15T23:59:59.000Z

    Precise predictions of the antineutrino spectra emitted by nuclear reactors is a key ingredient in measurements of reactor neutrino oscillations as well as in recent applications to the surveillance of power plants in the context of nonproliferation of nuclear weapons. We report new calculations including the latest information from nuclear databases and a detailed error budget. The first part of this work is the so-called ab initio approach where the total antineutrino spectrum is built from the sum of all {beta} branches of all fission products predicted by an evolution code. Systematic effects and missing information in nuclear databases lead to final relative uncertainties in the 10-20% range. A prediction of the antineutrino spectrum associated with the fission of {sup 238}U is given based on this ab initio method. For the dominant isotopes we developed a more accurate approach combining information from nuclear databases and reference electron spectra associated with the fission of {sup 235}U, {sup 239}Pu, and {sup 241}Pu, measured at Institut Laue-Langevin (ILL) in the 1980s. We show how the anchor point of the measured total {beta} spectra can be used to suppress the uncertainty in nuclear databases while taking advantage of all the information they contain. We provide new reference antineutrino spectra for {sup 235}U, {sup 239}Pu, and {sup 241}Pu isotopes in the 2-8 MeV range. While the shapes of the spectra and their uncertainties are comparable to those of the previous analysis of the ILL data, the normalization is shifted by about +3% on average. In the perspective of the reanalysis of past experiments and direct use of these results by upcoming oscillation experiments, we discuss the various sources of errors and their correlations as well as the corrections induced by off-equilibrium effects.

  4. Windscale pile reactors - Decommissioning progress on a fifty year legacy

    SciTech Connect (OSTI)

    Sexton, Richard J. [CH2M HILL International Nuclear Services, United Kingdom Atomic Energy Authority - UKAEA (United Kingdom)

    2007-07-01T23:59:59.000Z

    The decommissioning of the Windscale Pile 1 reactor, fifty years after the 1957 fire, is one of the most technically challenging decommissioning projects in the UK, if not the world. This paper presents a summary of the 1957 Windscale Pile 1 accident, its unique challenges and a new technical approach developed to safely and efficiently decommission the two Windscale Pile Reactors. The reactors will be decommissioned using a top down approach that employs an array of light weight, carbon fiber, high payload robotic arms to remove the damaged fuel, the graphite core, activated metals and concrete. This relatively conventional decommissioning approach has been made possible by a recently completed technical assessment of reactor core fire and criticality risk which concluded that these types of events are not credible if relatively simple controls are applied. This paper presents an overview of the design, manufacture and testing of equipment to remove the estimated 15 tons of fire damaged fuel and isotopes from the Pile 1 reactor. The paper also discusses recently conducted characterization activities which have allowed for a refined waste estimate and conditioning strategy. These data and an innovative approach have resulted in a significant reduction in the estimated project cost and schedule. (authors)

  5. Hypothetical Reactor Accident Study

    E-Print Network [OSTI]

    POPULATIONS; IODINE 131; MELTDOWN; METEOROLOGY; NUCLEAR POWER PLANTS; P CODES; PWR TYPE REACTORS; RADIATION in a Typical BWR and in a typical PWR. Comparison with WASH-1400 by C F . Højerup 202 APPENDIX 3. Calculation

  6. P Reactor Grouting

    SciTech Connect (OSTI)

    None

    2010-01-01T23:59:59.000Z

    Filling the P Reactor with grout. This seals the radioactive material and reduces the environmental footprint left from the Cold War. Project sponsored by the Recovery Act at the Savannah River Site.

  7. 08-G00333B_SNS_HFIR

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of ScienceandMesa del(ANL-IN-03-032) - Energy Innovation Portal AdvancedUsingLi Thermal1 from24

  8. Nuclear reactor control

    SciTech Connect (OSTI)

    Ingham, R.V.

    1980-01-01T23:59:59.000Z

    A liquid metal cooled fast breeder nuclear reactor has power setback means for use in an emergency. On initiation of a trip-signal a control rod is injected into the core in two stages, firstly, by free fall to effect an immediate power-set back to a safe level and, secondly, by controlled insertion. Total shut-down of the reactor under all emergencies is avoided. 4 claims.

  9. Polymerization reactor control

    SciTech Connect (OSTI)

    Ray, W.H.

    1985-01-01T23:59:59.000Z

    The principal difficulties in achieving good control of polymerization reactors are related to inadequate on-line measurement, a lack of understanding of the dynamics of the process, the highly sensitive and nonlinear behavior of these reactors, and the lack of well-developed techniques for the control of nonlinear processes. Some illustrations of these problems and a discussion of potential techniques for overcoming some of these difficulties is provided.

  10. Molten metal reactors

    DOE Patents [OSTI]

    Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

    2013-11-05T23:59:59.000Z

    A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

  11. Reactor- Nuclear Science Center 

    E-Print Network [OSTI]

    Unknown

    2011-08-17T23:59:59.000Z

    A COMPARISON OF NUCLEAR REACTOR CONTROL ROOM DISPLAY PANELS A Thesis by FRANCES RENAE BOWERS Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE May 1988... Major Subject: Industrial Engineering A COMPARISON OF NUCLEAR REACTOR CONTROL ROOM DISPLAY PANELS A Thesis by FRANCES RENAE BOWERS Approved as to style and content by: Rod er . oppa (Cha' of 'ttee) R. Quinn Brackett (Member) rome . Co gleton...

  12. RELAP5 Model of a Two-phase ThermoSyphon Experimental Facility for Fuels and Materials Irradiation

    SciTech Connect (OSTI)

    Carbajo, Juan J [ORNL] [ORNL; McDuffee, Joel Lee [ORNL] [ORNL

    2013-01-01T23:59:59.000Z

    The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) does not have a separate materials-irradiation flow loop and requires most materials and all fuel experiments to be placed inside a containment. This is necessary to ensure that internal contaminants such as fission products cannot be released into the primary coolant. As part of the safety basis justification, HFIR also requires that all experiments be able to withstand various accident conditions (e.g., loss of coolant) without generating vapor bubbles on the surface of the experiment in the primary coolant. As with any parallel flow system, HFIR is vulnerable to flow excursion events when vapor is generated in one of those flow paths. The effects of these requirements are to artificially increase experiment temperatures by introducing a barrier between the experimental materials and the HFIR coolant and to reduce experiment heat loads to ensure boiling doesn t occur. A new experimental facility for materials irradiation and testing in the HFIR is currently being developed to overcome these limitations. The new facility is unique in that it will have its own internal cooling flow totally independent of the reactor primary coolant and boiling is permitted. The reactor primary coolant will cool the outside of this facility without contacting the materials inside. The ThermoSyphon Test Loop (TSTL), a full scale prototype of the proposed irradiation facility to be tested outside the reactor, is being designed and fabricated (Ref. 1). The TSTL is a closed system working as a two-phase thermosyphon. A schematic is shown in Fig. 1. The bottom central part is the boiler/evaporator and contains three electric heaters. The vapor generated by the heaters will rise and be condensed in the upper condenser, the condensate will drain down the side walls and be circulated via a downcomer back into the bottom of the boiler. An external flow system provides coolant that simulates the HFIR primary coolant. The two-phase flow code RELAP5-3D (Ref. 2) is the main tool employed in this design. The model has multiple challenges: boiling, condensation and natural convection flows need to be modeled accurately.

  13. Power Burst Facility (PBF) Reactor Reactor Decommissioning

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible forPortsmouth/Paducah Project Office PressPostdoctoraldecadal7Powder DropperReactor

  14. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Dotson, CW

    1980-08-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  15. F Reactor Inspection

    SciTech Connect (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-10-29T23:59:59.000Z

    Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

  16. F Reactor Inspection

    ScienceCinema (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-11-24T23:59:59.000Z

    Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

  17. A Qualitative Assessment of Thorium-Based Fuels in Supercritical Pressure Water Cooled Reactors

    SciTech Connect (OSTI)

    Weaver, Kevan Dean; Mac Donald, Philip Elsworth

    2002-10-01T23:59:59.000Z

    The requirements for the next generation of reactors include better economics and safety, waste minimization (particularly of the long-lived isotopes), and better proliferation resistance (both intrinsic and extrinsic). A supercritical pressure water cooled reactor has been chosen as one of the lead contenders as a Generation IV reactor due to the high thermal efficiency and compact/simplified plant design. In addition, interest in the use of thorium-based fuels for Generation IV reactors has increased based on the abundance of thorium, and the minimization of transuranics in a neutron flux; as plutonium (and thus the minor actinides) is not a by-product in the thorium chain. In order to better understand the possibility of the combination of these concepts to meet the Generation IV goals, the qualitative burnup potential and discharge isotopics of thorium and uranium fuel were studied using pin cell analyses in a supercritical pressure water cooled reactor environment. Each of these fertile materials were used in both nitride and metallic form, with light water reactor grade plutonium and minor actinides added. While the uranium-based fuels achieved burnups that were 1.3 to 2.7 times greater than their thorium-based counterparts, the thorium-based fuels destroyed 2 to 7 times more of the plutonium and minor actinides. The fission-to-capture ratio is much higher in this reactor as compared to PWR’s and BWR’s due to the harder neutron spectrum, thus allowing more efficient destruction of the transuranic elements. However, while the uranium-based fuels do achieve a net depletion of plutonium and minor actinides, the breeding of these isotopes limits this depletion; especially as compared to the thorium-based fuels.

  18. Nickel isotopes in stellar matter

    E-Print Network [OSTI]

    Jameel-Un Nabi

    2014-08-19T23:59:59.000Z

    Isotopes of nickel play a key role during the silicon burning phase up to the presupernova phase of massive stars. Electron capture rates on these nickel isotopes are also important during the phase of core contraction. I present here the microscopic calculation of ground and excited states Gamow-Teller (GT) strength distributions for key nickel isotopes. The calculation is performed within the frame-work of pn-QRPA model. A judicious choice of model parameters, specially of the Gamow-Teller strength parameters and the deformation parameter, resulted in a much improved calculation of GT strength functions. The excited state GT distributions are much different from the corresponding ground-state distributions resulting in a failure of the Brink's hypothesis. The electron capture and positron decay rates on nickel isotopes are also calculated within the framework of pn-QRPA model relevant to the presupernova evolution of massive stars. The electron capture rates on odd-A isotopes of nickel are shown to have dominant contributions from parent excited states during as early as silicon burning phases. Comparison is being made with the large scale shell model calculation. During the silicon burning phases of massive stars the electron capture rates on $^{57,59}$Ni are around an order of magnitude bigger than shell model rates and can bear consequences for core-collapse simulators.

  19. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    SciTech Connect (OSTI)

    P. Delmolino

    2005-05-06T23:59:59.000Z

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  20. Mechanistic studies using kinetic isotope effects

    E-Print Network [OSTI]

    Schulmeier, Brian E.

    1999-01-01T23:59:59.000Z

    Understanding reaction mechanisms is an important aspect of chemistry. A now convenient way to study reaction mechanisms is kinetic isotope effects at natural abundance. This technique circumvents the cumbersome methods of traditional isotope effect...

  1. Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors

    DOE Patents [OSTI]

    Caldwell, John T. (Los Alamos, NM); Kunz, Walter E. (Santa Fe, NM); Atencio, James D. (Los Alamos, NM)

    1984-01-01T23:59:59.000Z

    A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify .sup.233 U, .sup.235 U and .sup.239 Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as .sup.240 Pu, .sup.244 Cm and .sup.252 Cf, and the spontaneous alpha particle emitter .sup.241 Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether "permanent" low-level burial is appropriate for the waste sample.

  2. Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors

    DOE Patents [OSTI]

    Caldwell, J.T.; Kunz, W.E.; Atencio, J.D.

    1982-03-31T23:59:59.000Z

    A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify /sup 233/U, /sup 235/U and /sup 239/Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as /sup 240/Pu, /sup 244/Cm and /sup 252/Cf, and the spontaneous alpha particle emitter /sup 241/Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether permanent low-level burial is appropriate for the waste sample.

  3. Zero Power Reactor simulation | Argonne National Laboratory

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

    Zero Power Reactor simulation Share Description Ever wanted to see a nuclear reactor core in action? Here's a detailed simulation of the Zero Power Reactor experiment, run by...

  4. Isotope Development & Production | Nuclear Science | ORNL

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

    Medical Radioisotope Radiochemical Separation & Processing Strategic Isotope Production Super Heavy Element Discovery Nuclear Security Science & Technology Nuclear Systems...

  5. Ultimate Isotope Precision for Carbonates Thermo Scientific

    E-Print Network [OSTI]

    Lachniet, Matthew S.

    Ultimate Isotope Precision for Carbonates Thermo Scientific KIEL IV Carbonate Device Part of Thermo integration cycle Ultimate Isotope Precision for Carbonates The Thermo Scientific KIEL IV Carbonate DeviceV Thermo Scientific MAT 253 or the 3-kV DELTA V isotope ratio mass spectrometer meets the requirements

  6. Isotope Cancer Treatment Research at LANL

    ScienceCinema (OSTI)

    Weidner, John; Nortier, Meiring

    2014-06-02T23:59:59.000Z

    Los Alamos National Laboratory has produced medical isotopes for diagnostic and imaging purposes for more than 30 years. Now LANL researchers have branched out into isotope cancer treatment studies. New results show that an accelerator-based approach can produce clinical trial quantities of actinium-225, an isotope that has promise as a way to kill tumors without damaging surrounding healthy cells.

  7. Methanation assembly using multiple reactors

    DOE Patents [OSTI]

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24T23:59:59.000Z

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  8. U.S. Plans for the Next Fast Reactor Transmutation Fuels Irradiation Test

    SciTech Connect (OSTI)

    B. A. Hilton

    2007-09-01T23:59:59.000Z

    The U.S. Advanced Fuel Cycle Initiative (AFCI) seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products, thereby dramatically decreasing the volume of material requiring disposal and the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository. One important component of the technology development is actinide-bearing transmutation fuel forms containing plutonium, neptunium, americium (and possibly curium) isotopes. Metallic alloy and oxide fuel forms are being developed as the near term options for fast reactor implementation.

  9. Assessment of Non-traditional Isotopic Ratios by Mass Spectrometry for Analysis of Nuclear Activities: Annual Report Year 2

    SciTech Connect (OSTI)

    Biegalski, S; Buchholz, B

    2009-08-26T23:59:59.000Z

    The objective of this work is to identify isotopic ratios suitable for analysis via mass spectrometry that distinguish between commercial nuclear reactor fuel cycles, fuel cycles for weapons grade plutonium, and products from nuclear weapons explosions. Methods will also be determined to distinguish the above from medical and industrial radionuclide sources. Mass spectrometry systems will be identified that are suitable for field measurement of such isotopes in an expedient manner. Significant progress has been made with this project within the past year: (1) Isotope production from commercial nuclear fuel cycles and nuclear weapons fuel cycles have been modeled with the ORIGEN and MCNPX codes. (2) MCNPX has been utilized to calculate isotopic inventories produced in a short burst fast bare sphere reactor (to approximate the signature of a nuclear weapon). (3) Isotopic ratios have been identified that are good for distinguishing between commercial and military fuel cycles as well as between nuclear weapons and commercial nuclear fuel cycles. (4) Mass spectrometry systems have been assessed for analysis of the fission products of interest. (5) A short-list of forensic ratios have been identified that are well suited for use in portable mass spectrometry systems.

  10. Light water reactor mixed-oxide fuel irradiation experiment

    SciTech Connect (OSTI)

    Hodge, S.A.; Cowell, B.S. [Oak Ridge National Lab., TN (United States); Chang, G.S.; Ryskamp, J.M. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.

    1998-06-01T23:59:59.000Z

    The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding.

  11. A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor...

    Office of Scientific and Technical Information (OSTI)

    A reactor and associated power plant designed to produce 1.05 Mwh and 3.535 Mwh of steam for heating purposes are described. The total thermal output of the reactor is 10 Mwh....

  12. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect (OSTI)

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01T23:59:59.000Z

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  13. Application of advanced liquid metal reactors to the destruction of radioactive wastes

    SciTech Connect (OSTI)

    Karnesky, R.A.; Dobbin, K.D.; Jordheim, D.P.; Rawlins, J.A.; Wootan, D.W.

    1991-08-01T23:59:59.000Z

    The ability of a small fast reactor to destroy hazardous long lived minor actinide and fission product wastes is evaluated. It is determined that by using a novel technique wherein high energy neutrons leaking from the active core of the reactor are moderated by yttrium hydride located in the target assemblies, substantial amounts of long lived fission products can be destroyed and useful quantities of the beneficial isotope {sup 238}Pu can be produced by transmutation of the {sup 237}Np and {sub 241}Am minor actinide waste components. In addition it is shown that minor actinides recovered from spent Light Water Reactor fuel can be used to fuel such a reactor, increasing the amount of hazardous minor actinide and fission product wastes that can be destroyed. 9 refs., 4 figs., 2 tabs.

  14. Uncertainties analysis of fission fraction for reactor antineutrino experiments using DRAGON

    E-Print Network [OSTI]

    X. B. Ma; L. Z. Wang; Y. X. Chen; W. L. Zhong; F. P. An

    2014-05-27T23:59:59.000Z

    Rising interest in nuclear reactors as a source of antineutrinos for experiments motivates validated, fast, and accessible simulation to predict reactor rates. First, DRAGON was developed to calculate the fission rates of the four most important isotopes in fissions,235U,238U,239Pu and141Pu, and it was validated for PWRs using the Takahama benchmark. The fission fraction calculation function was validated through comparing our calculation results with MIT's results. we calculate the fission fraction of the Daya Bay reactor core, and compare its with those calculated by the commercial reactor simulation program SCIENCE, which is used by the Daya Bay nuclear power plant, and the results was consist with each other. The uncertainty of the antineutrino flux by the fission fraction was studied, and the uncertainty of the antineutrino flux by the fission fraction simulation is 0.6% per core for Daya Bay antineutrino experiment.

  15. Reactor for exothermic reactions

    DOE Patents [OSTI]

    Smith, Jr., Lawrence A. (Bellaire, TX); Hearn, Dennis (Houston, TX); Jones, Jr., Edward M. (Friendswood, TX)

    1993-01-01T23:59:59.000Z

    A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120.degree. to 300.degree. F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  16. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-08-01T23:59:59.000Z

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  17. Reactor for exothermic reactions

    DOE Patents [OSTI]

    Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

    1993-03-02T23:59:59.000Z

    A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  18. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, Ernest (Wilmette, IL)

    1986-01-01T23:59:59.000Z

    A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

  19. Fusion reactor control

    SciTech Connect (OSTI)

    Plummer, D.A.

    1995-12-31T23:59:59.000Z

    The plasma kinetic temperature and density changes, each per an injected fuel density rate increment, control the energy supplied by a thermonuclear fusion reactor in a power production cycle. This could include simultaneously coupled control objectives for plasma current, horizontal and vertical position, shape and burn control. The minimum number of measurements required, use of indirect (not plasma parameters) system measurements, and distributed control procedures for burn control are to be verifiable in a time dependent systems code. The International Thermonuclear Experimental Reactor (ITER) has the need to feedback control both the fusion output power and the driven plasma current, while avoiding damage to diverter plates. The system engineering of fusion reactors must be performed to assure their development expeditiously and effectively by considering reliability, availability, maintainability, environmental impact, health and safety, and cost.

  20. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

    1987-01-01T23:59:59.000Z

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

  1. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21T23:59:59.000Z

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  2. Fusion reactor high vacuum pumping: Charcoal cryosorber tritium exposure results

    SciTech Connect (OSTI)

    Sedgley, D.W.; Walthers, C.R.; Jenkins, E.M. (Grumman Aerospace Corp., Bethpage, NY (United States))

    1991-01-01T23:59:59.000Z

    Recent experiments, have shown the practically of using activated charcoal (coconut charcoal) at 4{degrees}K to pump helium and hydrogen isotopes for a fusion reactor. Both speed and capacity for deuterium/helium and tritium/helium-3 mixtures were shown to be satisfactory. The long term effects of tritium on the charcoal/cement system developed by Grumman and LLNL were not known and a program was undertaken to see what, if any, effect long term tritium exposure has on the cryosorber. Several charcoal on aluminum test samples were subjected to six months exposure of tritium at approximately 77{degrees}K. The tritium was scanned several times with a residual gas analyzer and the speed-capacity performance of the samples was measured before, approximately half way through and after the exposure. Modest effects were noted which would not seriously restrict charcoal's use as a cryosorber for fusion reactor high vacuum pumping applications. 4 refs., 8 figs.

  3. Retrospective dosimetry analyses of reactor vessel cladding samples

    SciTech Connect (OSTI)

    Greenwood, L. R.; Soderquist, C. Z. [Battelle Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Fero, A. H. [Westinghouse Electric Company, Cranberry Twp., PA 16066 (United States)

    2011-07-01T23:59:59.000Z

    Reactor pressure vessel cladding samples for Ringhals Units 3 and 4 in Sweden were analyzed using retrospective reactor dosimetry techniques. The objective was to provide the best estimates of the neutron fluence for comparison with neutron transport calculations. A total of 51 stainless steel samples consisting of chips weighing approximately 100 to 200 mg were removed from selected locations around the pressure vessel and were sent to Pacific Northwest National Laboratory for analysis. The samples were fully characterized and analyzed for radioactive isotopes, with special interest in the presence of Nb-93m. The RPV cladding retrospective dosimetry results will be combined with a re-evaluation of the surveillance capsule dosimetry and with ex-vessel neutron dosimetry results to form a comprehensive 3D comparison of measurements to calculations performed with 3D deterministic transport code. (authors)

  4. Fast quench reactor method

    DOE Patents [OSTI]

    Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID); Berry, Ray A. (Idaho Falls, ID)

    1999-01-01T23:59:59.000Z

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

  5. Fast quench reactor method

    DOE Patents [OSTI]

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.; Berry, R.A.

    1999-08-10T23:59:59.000Z

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream. 8 figs.

  6. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, Daniel L. (Princeton, NJ)

    1988-01-01T23:59:59.000Z

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.

  7. Reactor Neutrino Experiments

    E-Print Network [OSTI]

    Jun Cao

    2007-12-06T23:59:59.000Z

    Precisely measuring $\\theta_{13}$ is one of the highest priority in neutrino oscillation study. Reactor experiments can cleanly determine $\\theta_{13}$. Past reactor neutrino experiments are reviewed and status of next precision $\\theta_{13}$ experiments are presented. Daya Bay is designed to measure $\\sin^22\\theta_{13}$ to better than 0.01 and Double Chooz and RENO are designed to measure it to 0.02-0.03. All are heading to full operation in 2010. Recent improvements in neutrino moment measurement are also briefed.

  8. Perspectives on reactor safety

    SciTech Connect (OSTI)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01T23:59:59.000Z

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  9. THE MATERIALS OF FAST BREEDER REACTORS

    E-Print Network [OSTI]

    Olander, Donald R.

    2013-01-01T23:59:59.000Z

    times larger in a fast reactor than in a thermal reactor,structural metals in a fast reactor will be subject to farof fuel ele- ments in fast reactors which are roughly one

  10. Innovative design of uranium startup fast reactors

    E-Print Network [OSTI]

    Fei, Tingzhou

    2012-01-01T23:59:59.000Z

    Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

  11. Reactor operation environmental information document

    SciTech Connect (OSTI)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01T23:59:59.000Z

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  12. Power calibrations for TRIGA reactors

    SciTech Connect (OSTI)

    Whittemore, W.L.; Razvi, J.; Shoptaugh, J.R. Jr. [General Atomics, San Diego, CA (United States)

    1988-07-01T23:59:59.000Z

    The purpose of this paper is to establish a framework for the calorimetric power calibration of TRIGA reactors so that reliable results can be obtained with a precision better than {+-} 5%. Careful application of the same procedures has produced power calibration results that have been reproducible to {+-} 1.5%. The procedures are equally applicable to the Mark I, Mark II and Mark III reactors as well as to reactors having much larger reactor tanks and to TRIGA reactors capable of forced cooling up to 3 MW in some cases and 15 MW in another case. In the case of forced cooled TRIGA reactors, the calorimetric power calibration is applicable in the natural convection mode for these reactors using exactly the same procedures as are discussed below for the smaller TRIGA reactors (< 2 MW)

  13. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    SciTech Connect (OSTI)

    B. Boer; A. M. Ougouag

    2010-09-01T23:59:59.000Z

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).

  14. Reactor operation safety information document

    SciTech Connect (OSTI)

    Not Available

    1990-01-01T23:59:59.000Z

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  15. Reed Reactor Facility Annual Report

    SciTech Connect (OSTI)

    Frantz, Stephen G.

    2000-09-01T23:59:59.000Z

    This is the report of the operations, experiments, modifications, and other aspects of the Reed Reactor Facility for the year.

  16. Feasibility study Part I - Thermal hydraulic analysis of LEU target for {sup 99}Mo production in Tajoura reactor

    SciTech Connect (OSTI)

    Bsebsu, F.M.; Abotweirat, F. [Reactor Department, Renewable Energies and Water Desalination Research Cente, P.O. Box 30878 Tajoura, Tripoli (Libyan Arab Jamahiriya)], E-mail: Bsebso@yahoo.com, E-mail: abutweirat@yahoo.com; Elwaer, S. [Radiochemistry Department, Renewable Energies and Water Desalination Research Cente, P.O. Box 30878 Tajoura, Tripoli (Libyan Arab Jamahiriya)], E-mail: samiwer@yahoo.com

    2008-07-15T23:59:59.000Z

    The Renewable Energies and Water Desalination Research Center (REWDRC), Libya, will implement the technology for {sup 99}Mo isotope production using LEU foil target, to obtain new revenue streams for the Tajoura nuclear research reactor and desiring to serve the Libyan hospitals by providing the medical radioisotopes. Design information is presented for LEU target with irradiation device and irradiation Beryllium (Be) unit in the Tajoura reactor core. Calculated results for the reactor core with LEU target at different level of power are presented for steady state and several reactivity induced accident situations. This paper will present the steady state thermal hydraulic design and transient analysis of Tajoura reactor was loaded with LEU foil target for {sup 99}Mo production. The results of these calculations show that the reactor with LEU target during the several cases of transient are in safe and no problems will occur. (author)

  17. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    SciTech Connect (OSTI)

    Walter G. Luscher; David J. Senor; Keven K. Clayton; Glen R. Longhurst

    2015-01-01T23:59:59.000Z

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

  18. In-Reactor Oxidation of Zircaloy-4 Under Low Water Vapor Pressures

    SciTech Connect (OSTI)

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin; Longhurst, Glen

    2015-01-01T23:59:59.000Z

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330° and 370°C). Data from these tests will be used to support fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex- reactor test results were performed to evaluate the influence of irradiation.

  19. Nuclear Reactors and Technology

    SciTech Connect (OSTI)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01T23:59:59.000Z

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  20. NETL - Chemical Looping Reactor

    ScienceCinema (OSTI)

    None

    2014-06-26T23:59:59.000Z

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  1. Thermal Reactor Safety

    SciTech Connect (OSTI)

    Not Available

    1980-06-01T23:59:59.000Z

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  2. Nuclear reactor building

    DOE Patents [OSTI]

    Gou, Perng-Fei (Saratoga, CA); Townsend, Harold E. (Campbell, CA); Barbanti, Giancarlo (Sirtori, IT)

    1994-01-01T23:59:59.000Z

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  3. Nuclear reactor building

    DOE Patents [OSTI]

    Gou, P.F.; Townsend, H.E.; Barbanti, G.

    1994-04-05T23:59:59.000Z

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

  4. Fossil fuel furnace reactor

    DOE Patents [OSTI]

    Parkinson, William J. (Los Alamos, NM)

    1987-01-01T23:59:59.000Z

    A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

  5. NETL - Chemical Looping Reactor

    SciTech Connect (OSTI)

    None

    2013-07-24T23:59:59.000Z

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  6. Reactor Monitoring with Neutrinos

    E-Print Network [OSTI]

    M. Cribier

    2007-04-06T23:59:59.000Z

    The fundamental knowledge on neutrinos acquired in the recent years open the possibility of applied neutrino physics. Among it the automatic and non intrusive monitoring of nuclear reactor by its antineutrino signal could be very valuable to IAEA in charge of the control of nuclear power plants. Several efforts worldwide have already started.

  7. Gamma Spectrum from Neutron Capture on Tungsten Isotopes

    SciTech Connect (OSTI)

    Hurst, Aaron; Summers, Neil; Sleaford, Brad; Firestone, Richard B; Belgya, T.; Revay, Z.S.

    2010-04-29T23:59:59.000Z

    An evaluation of thermal neutron capture on the stable tungsten isotopes is presented, with preliminary results for the compound systems 183;184;185;187W. The evaluation procedure compares the g-ray cross-section data collected at the Budapest reactor, with Monte Carlo simulations of g-ray emission following the thermal neutron-capture process. The statistical-decay code DICEBOX was used for the Monte Carlo simulations. The evaluation yields new gamma rays in 185W and the confirmation of spins in 187W, raising the number of levels below which the level schemes are considered complete, thus increasing the number of levels that can be used in neutron data libraries.

  8. PROSPECT - A precision oscillation and spectrum experiment

    E-Print Network [OSTI]

    ,

    2015-01-01T23:59:59.000Z

    Segmented antineutrino detectors placed near a compact research reactor provide an excellent opportunity to probe short-baseline neutrino oscillations and precisely measure the reactor antineutrino spectrum. Close proximity to a reactor combined with minimal overburden yield a high background environment that must be managed through shielding and detector technology. PROSPECT is a new experimental effort to detect reactor antineutrinos from the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory, managed by UT Battelle for the U.S. Department of Energy. The detector will use novel lithium-loaded liquid scintillator capable of neutron/gamma pulse shape discrimination and neutron capture tagging. These enhancements improve the ability to identify neutrino inverse-beta decays and reject background events in analysis. Results from these efforts will be covered along with their implications for an oscillation search and a precision spectrum measurement.

  9. PROSPECT - A precision oscillation and spectrum experiment

    E-Print Network [OSTI]

    T. J. Langford

    2014-12-22T23:59:59.000Z

    Segmented antineutrino detectors placed near a compact research reactor provide an excellent opportunity to probe short-baseline neutrino oscillations and precisely measure the reactor antineutrino spectrum. Close proximity to a reactor combined with minimal overburden yield a high background environment that must be managed through shielding and detector technology. PROSPECT is a new experimental effort to detect reactor antineutrinos from the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory, managed by UT Battelle for the U.S. Department of Energy. The detector will use novel lithium-loaded liquid scintillator capable of neutron/gamma pulse shape discrimination and neutron capture tagging. These enhancements improve the ability to identify neutrino inverse-beta decays and reject background events in analysis. Results from these efforts will be covered along with their implications for an oscillation search and a precision spectrum measurement.

  10. Metal-fueled HWR (heavy water reactors) severe accident issues: Differences and similarities to commercial LWRs (light water reactors)

    SciTech Connect (OSTI)

    Ellison, P.G.; Hyder, M.L.; Monson, P.R. (Westinghouse Savannah River Co., Aiken, SC (USA)); Coryell, E.W. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-01-01T23:59:59.000Z

    Differences and similarities in severe accident progression and phenomena between commercial Light Water Reactors (LWR) and metal-fueled isotopic production Heavy Water Reactors (HWR) are described. It is very important to distinguish between accident progression in the two systems because each reactor type behaves in a unique manner to a fuel melting accident. Some of the lessons learned as a result of the extensive commercial severe accident research are not applicable to metal-fueled heavy water reactors. A direct application of severe accident phenomena developed from oxide-fueled LWRs to metal-fueled HWRs may lead to large errors or substantial uncertainties. In general, the application of severe accident LWR concepts to HWRs should be done with the intent to define the relevant issues, define differences, and determine areas of overlap. This paper describes the relevant differences between LWR and metal-fueled HWR severe accident phenomena. Also included in the paper is a description of the phenomena that govern the source term in HWRs, the areas where research is needed to resolve major uncertainties, and areas in which LWR technology can be directly applied with few modifications.

  11. Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2006-10-01T23:59:59.000Z

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  12. Reactor vessel support system. [LMFBR

    DOE Patents [OSTI]

    Golden, M.P.; Holley, J.C.

    1980-05-09T23:59:59.000Z

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  13. Implications of Fast Reactor Transuranic Conversion Ratio

    SciTech Connect (OSTI)

    Steven J. Piet; Edward A. Hoffman; Samuel E. Bays

    2010-11-01T23:59:59.000Z

    Theoretically, the transuranic conversion ratio (CR), i.e. the transuranic production divided by transuranic destruction, in a fast reactor can range from near zero to about 1.9, which is the average neutron yield from Pu239 minus 1. In practice, the possible range will be somewhat less. We have studied the implications of transuranic conversion ratio of 0.0 to 1.7 using the fresh and discharge fuel compositions calculated elsewhere. The corresponding fissile breeding ratio ranges from 0.2 to 1.6. The cases below CR=1 (“burners”) do not have blankets; the cases above CR=1 (“breeders”) have breeding blankets. The burnup was allowed to float while holding the maximum fluence to the cladding constant. We graph the fuel burnup and composition change. As a function of transuranic conversion ratio, we calculate and graph the heat, gamma, and neutron emission of fresh fuel; whether the material is “attractive” for direct weapon use using published criteria; the uranium utilization and rate of consumption of natural uranium; and the long-term radiotoxicity after fuel discharge. For context, other cases and analyses are included, primarily once-through light water reactor (LWR) uranium oxide fuel at 51 MWth-day/kg-iHM burnup (UOX-51). For CR<1, the heat, gamma, and neutron emission increase as material is recycled. The uranium utilization is at or below 1%, just as it is in thermal reactors as both types of reactors require continuing fissile support. For CR>1, heat, gamma, and neutron emission decrease with recycling. The uranium utilization exceeds 1%, especially as all the transuranic elements are recycled. exceeds 1%, especially as all the transuranic elements are recycled. At the system equilibrium, heat and gamma vary by somewhat over an order of magnitude as a function of CR. Isotopes that dominate heat and gamma emission are scattered throughout the actinide chain, so the modest impact of CR is unsurprising. Neutron emitters are preferentially found among the higher actinides, so the neutron emission varies much stronger with CR, about three orders of magnitude.

  14. PROSPECT - A Precision Reactor Oscillation and Spectrum Experiment at Short Baselines

    E-Print Network [OSTI]

    J. Ashenfelter; A. B. Balantekin; H. R. Band; G. Barclay; C. Bass; N. S. Bowden; C. D. Bryan; J. J. Cherwinka; R. Chu; T. Classen; D. Davee; D. Dean; G. Deichert; M. Diwan; M. J. Dolinski; J. Dolph; D. A. Dwyer; Y. Efremenko; S. Fan; A. Galindo-Uribarri; K. Gilje; A. Glenn; M. Green; K. Han; S. Hans; K. M. Heeger; B. Heffron; L. Hu; P. Huber; D. E. Jaffe; Y. Kamyshkov; S. Kettell; C. Lane; T. J. Langford; B. R. Littlejohn; D. Martinez; R. D. McKeown; M. P. Mendenhall; S. Morrell; P. Mueller; H. P. Mumm; J. Napolitano; J. S. Nico; D. Norcini; D. Pushin; X. Qian; E. Romero; R. Rosero; B. S. Seilhan; R. Sharma; P. T. Surukuchi; S. J. Thompson; R. L. Varner; B. Viren; W. Wang; B. White; C. White; J. Wilhelmi; C. Williams; R. E. Williams; T. Wise; H. Yao; M. Yeh; N. Zaitseva; C. Zhang; X. Zhang

    2015-01-27T23:59:59.000Z

    Current models of antineutrino production in nuclear reactors predict detection rates and spectra at odds with the existing body of direct reactor antineutrino measurements. High-resolution antineutrino detectors operated close to compact research reactor cores can produce new precision measurements useful in testing explanations for these observed discrepancies involving underlying nuclear or new physics. Absolute measurement of the 235U-produced antineutrino spectrum can provide additional constraints for evaluating the accuracy of current and future reactor models, while relative measurements of spectral distortion between differing baselines can be used to search for oscillations arising from the existence of eV-scale sterile neutrinos. Such a measurement can be performed in the United States at several highly-enriched uranium fueled research reactors using near-surface segmented liquid scintillator detectors. We describe here the conceptual design and physics potential of the PROSPECT experiment, a U.S.-based, multi-phase experiment with reactor-detector baselines of 7-20 meters capable of addressing these and other physics and detector development goals. Current R&D status and future plans for PROSPECT detector deployment and data-taking at the High Flux Isotope Reactor at Oak Ridge National Laboratory will be discussed.

  15. Observations of Fallout from the Fukushima Reactor Accident in San Francisco Bay Area Rainwater

    E-Print Network [OSTI]

    Eric B. Norman; Christopher T. Angell; Perry A. Chodash

    2011-03-30T23:59:59.000Z

    We have observed fallout from the recent Fukushima Dai-ichi reactor accident in samples of rainwater collected in the San Francisco Bay area. Gamma ray spectra measured from these samples show clear evidence of fission products - 131,132I, 132Te, and 134,137Cs. The activity levels we have measured for these isotopes are very low and pose no health risk to the public.

  16. Nuclear reactor construction with bottom supported reactor vessel

    DOE Patents [OSTI]

    Sharbaugh, John E. (Bullskin Township, Fayette County, PA)

    1987-01-01T23:59:59.000Z

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

  17. Validation Work to Support the Idaho National Engineering and Environmental Laboratory Calculational Burnup Methodology Using Shippingport Light Water Breeder Reactor (LWBR) Spent Fuel Assay Data

    SciTech Connect (OSTI)

    J. W. Sterbentz

    1999-08-01T23:59:59.000Z

    Six uranium isotopes and fourteen fission product isotopes were calculated on a mass basis at end-of-life (EOL) conditions for three fuel rods from different Light Water Breeder Reactor (LWBR) measurements. The three fuel rods evaluated here were taken from an LWBR seed module, a standard blanket module, and a reflector (Type IV) module. The calculated results were derived using a depletion methodology previously employed to evaluate many of the radionuclide inventories for spent nuclear fuels at the Idaho National Engineering and Environmental Laboratory. The primary goal of the calculational task was to further support the validation of this particular calculational methodology and its application to diverse reactor types and fuels. Result comparisons between the calculated and measured mass concentrations in the three rods indicate good agreement for the three major uranium isotopes (U-233, U-234, U-235) with differences of less than 20%. For the seed and standard blanket rod, the U-233 and U-234 differences were within 5% of the measured values (these two isotopes alone represent greater than 97% of the EOL total uranium mass). For the major krypton and xenon fission product isotopes, differences of less than 20% and less than 30% were observed, respectively. In general, good agreement was obtained for nearly all the measured isotopes. For these isotopes exhibiting significant differences, possible explanations are discussed in terms of measurement uncertainty, complex transmutations, etc.

  18. Analysis of hydrogen isotope mixtures

    DOE Patents [OSTI]

    Villa-Aleman, Eliel (Aiken, SC)

    1994-01-01T23:59:59.000Z

    An apparatus and method for determining the concentrations of hydrogen isotopes in a sample. Hydrogen in the sample is separated from other elements using a filter selectively permeable to hydrogen. Then the hydrogen is condensed onto a cold finger or cryopump. The cold finger is rotated as pulsed laser energy vaporizes a portion of the condensed hydrogen, forming a packet of molecular hydrogen. The desorbed hydrogen is ionized and admitted into a mass spectrometer for analysis.

  19. Surface radiological investigations at the proposed SWSA 7 Site, Oak Ridge National Laboratory, Oak Ridge, Tennessee

    SciTech Connect (OSTI)

    McKenzie, S.P.; Murray, M.E.; Uziel, M.S.

    1995-08-01T23:59:59.000Z

    A surface radiological investigation was conducted intermittently from June 1994 to June 1995 at the proposed site for Solid Waste Storage Area (SWSA) 7. The stimulus for this survey was the observation in June 1992 of a man`s trousers became contaminated with {sup 9O}Sr while he was reviewing work on top of the High Flux Isotope Reactor (HFIR) cooling tower. Radiation surveys identified {sup 9O}Sr on the roofs of older buildings at the HFIR site. Since no {sup 9O}Sr was found on buildings built between 1988 and 1990, the {sup 9O}Sr was thought to have been deposited prior to 1988. Later in 1992, beta particles were identified on a bulldozer that had been used in a wooded area southwest of the Health Physics Research Reactor (HPRR) Access Road. More recently in April 1995, {sup 9O}Sr particles were identified on the top side of ceiling tiles in the overhead area of a building in the HFIR Complex. Considering that the proposed SWSA 7 site was located between the HFIR complex and the HPRR Access Road, it was deemed prudent to investigate the possibility that beta particles might also be present at the SWSA 7 site. A possible explanation for the presence of these particles has been provided by long-time ORNL employees and retirees. Strontium-90 as the titanate was developed in the early 1960s as part of the Systems for Nuclear Auxiliary Power (SNAP) Program. Strontium titanate ({sup 90}SrTiO{sub 3}) was produced at the Fission Product Development Laboratory (Building 3517) in the ORNL main plant area. Waste from the process was loaded into a 1-in. lead-lined dumpster, which was transferred to SWSA 5 where it was dumped into a trench. Dumping allowed some articles to become airborne.

  20. Stable Isotope Signatures for Microbial Forensics

    SciTech Connect (OSTI)

    Kreuzer, Helen W.

    2012-01-03T23:59:59.000Z

    The isotopic distribution of the atoms composing the molecules of microorganisms is a function of the substrates used by the organisms. The stable isotope content of an organism is fixed so long as no further substrate consumption and biosynthesis occurs, while the radioactive isotopic content decays over time. The distribution of stable isotopes of C, N, O and H in heterotrophic microorganisms is a direct function of the culture medium, and therefore the stable isotope composition can be used to associate samples with potential culture media and also with one another. The 14C content depends upon the 14C content, and therefore the age, of the organic components of the culture medium, as well as on the age of the culture itself. Stable isotope signatures can thus be used for sample matching, to associate cultures with specific growth media, and to predict characteristics of growth media.

  1. Stable Isotope Enrichment Capabilities at ORNL

    SciTech Connect (OSTI)

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

    2013-01-01T23:59:59.000Z

    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.

  2. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, E.

    1983-08-15T23:59:59.000Z

    A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

  3. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, D.L.

    1987-09-04T23:59:59.000Z

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

  4. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Peng, Yueng-Kay M. (Oak Ridge, TN)

    1989-01-01T23:59:59.000Z

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  5. Nuclear reactor sealing system

    DOE Patents [OSTI]

    McEdwards, James A. (Calabasas, CA)

    1983-01-01T23:59:59.000Z

    A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

  6. Stable isotope signatures for characterising the biological stability of landfilled municipal solid waste

    SciTech Connect (OSTI)

    Wimmer, Bernhard, E-mail: bernhard.wimmer@ait.ac.at [AIT Austrian Institute of Technology GmbH, Health and Environment Department, Environmental Resources and Technologies, Konrad-Lorenz-Strasse 24, 3430 Tulln (Austria); Hrad, Marlies; Huber-Humer, Marion [Institute of Waste Management, Department of Water-Atmosphere-Environment, University of Natural Resources and Life Sciences, Muthgasse 107, 1190 Vienna (Austria); Watzinger, Andrea; Wyhlidal, Stefan; Reichenauer, Thomas G. [AIT Austrian Institute of Technology GmbH, Health and Environment Department, Environmental Resources and Technologies, Konrad-Lorenz-Strasse 24, 3430 Tulln (Austria)

    2013-10-15T23:59:59.000Z

    Highlights: ? The isotopic signature of ?{sup 13}C-DIC of leachates is linked to the reactivity of MSW. ? Isotopic signatures of leachates depend on aerobic/anaerobic conditions in landfills. ? In situ aeration of landfills can be monitored by isotope analysis in leachate. ? The isotopic analysis of leachates can be used for assessing the stability of MSW. ? ?{sup 13}C-DIC of leachates helps to define the duration of landfill aftercare. - Abstract: Stable isotopic signatures of landfill leachates are influenced by processes within municipal solid waste (MSW) landfills mainly depending on the aerobic/anaerobic phase of the landfill. We investigated the isotopic signatures of ?{sup 13}C, ?{sup 2}H and ?{sup 18}O of different leachates from lab-scale experiments, lysimeter experiments and a landfill under in situ aeration. In the laboratory, columns filled with MSW of different age and reactivity were percolated under aerobic and anaerobic conditions. In landfill simulation reactors, waste of a 25 year old landfill was kept under aerobic and anaerobic conditions. The lysimeter facility was filled with mechanically shredded fresh waste. After starting of the methane production the waste in the lysimeter containments was aerated in situ. Leachate and gas composition were monitored continuously. In addition the seepage water of an old landfill was collected and analysed periodically before and during an in situ aeration. We found significant differences in the ?{sup 13}C-value of the dissolved inorganic carbon (?{sup 13}C-DIC) of the leachate between aerobic and anaerobic waste material. During aerobic degradation, the signature of ?{sup 13}C-DIC was mainly dependent on the isotopic composition of the organic matter in the waste, resulting in a ?{sup 13}C-DIC of ?20‰ to ?25‰. The production of methane under anaerobic conditions caused an increase in ?{sup 13}C-DIC up to values of +10‰ and higher depending on the actual reactivity of the MSW. During aeration of a landfill the aerobic degradation of the remaining organic matter caused a decrease to a ?{sup 13}C-DIC of about ?20‰. Therefore carbon isotope analysis in leachates and groundwater can be used for tracing the oxidation–reduction status of MSW landfills. Our results indicate that monitoring of stable isotopic signatures of landfill leachates over a longer time period (e.g. during in situ aeration) is a powerful and cost-effective tool for characterising the biodegradability and stability of the organic matter in landfilled municipal solid waste and can be used for monitoring the progress of in situ aeration.

  7. Stable Isotope Protocols: Sampling and Sample Processing

    E-Print Network [OSTI]

    Levin, Lisa A; Currin, Carolyn

    2012-01-01T23:59:59.000Z

    plants, benthic microalgae [BMI], benthic macroalgae) andin a dessicator, prior to analysis. A.2 Benthic microalgaeBenthic microalgae (BMI) can be collected for isotope

  8. EIS-0249: Medical Isotopes Production Project

    Broader source: Energy.gov [DOE]

    This EIS evaluates the potential environmental impacts of a proposal to establish a production capability for molybdenum-99 (Mo-99) and related medical isotopes.

  9. Integration of Nontraditional Isotopic Systems Into Reaction...

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

    Water-Rock Interaction, and Impacts of Water Chemistry on Reservoir Sustainability Integration of Nontraditional Isotopic Systems Into Reaction-Transport Models of EGS For...

  10. Atom trap trace analysis of krypton isotopes

    SciTech Connect (OSTI)

    Bailey, K.; Chen, C. Y.; Du, X.; Li, Y. M.; Lu, Z.-T.; O'Connor, T. P.; Young, L.

    1999-11-17T23:59:59.000Z

    A new method of ultrasensitive isotope trace analysis has been developed. This method, based on the technique of laser manipulation of neutral atoms, has been used to count individual {sup 85}Kr and {sup 81}Kr atoms present in a natural krypton gas sample with isotopic abundances in the range of 10{sup {minus}11} and 10{sup {minus}13}, respectively. This method is free of contamination from other isotopes and elements and can be applied to several different isotope tracers for a wide range of applications. The demonstrated detection efficiency is 1 x 10{sup {minus}7}. System improvements could increase the efficiency by many orders of magnitude.

  11. Stable Isotope Fractionations in Biogeochemical Reactive Transport

    E-Print Network [OSTI]

    Druhan, Jennifer Lea

    2012-01-01T23:59:59.000Z

    34 S fractionation . Summary A mesoscale study of isotopicion exchange and ! 44 Ca . A mesoscale study of isotopicmodeling and ! 34 S . A mesoscale study of isotopic

  12. Method for isotope enrichment by photoinduced chemiionization

    DOE Patents [OSTI]

    Dubrin, James W. (Walnut Creek, CA)

    1985-01-01T23:59:59.000Z

    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.

  13. O isotopic composition of CaCO3 measured by continuous ow isotope ratio mass spectrometry

    E-Print Network [OSTI]

    d13 C and d18 O isotopic composition of CaCO3 measured by continuous ¯ow isotope ratio mass. This new method streamlines the classical phosphoric acid/calcium carbonate (H3PO4/CaCO3) reaction method XL continuous flow isotope ratio mass spectrometer. Conditions for which the H3PO4/CaCO3 reaction

  14. SAVANNAH RIVER SITE R REACTOR DISASSEMBLY BASIN IN SITU DECOMMISSIONING

    SciTech Connect (OSTI)

    Langton, C.; Blankenship, J.; Griffin, W.; Serrato, M.

    2009-12-03T23:59:59.000Z

    The US DOE concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate in tact, structurally sound facilities that are no longer needed for their original purpose of, i.e., generating (reactor facilities), processing(isotope separation facilities) or storing radioactive materials. The 105-R Disassembly Basin is the first SRS reactor facility to undergo the in-situ decommissioning (ISD) process. This ISD process complies with the105-R Disassembly Basin project strategy as outlined in the Engineering Evaluation/Cost Analysis for the Grouting of the R-Reactor Disassembly Basin at the Savannah River Site and includes: (1) Managing residual water by solidification in-place or evaporation at another facility; (2) Filling the below grade portion of the basin with cementitious materials to physically stabilize the basin and prevent collapse of the final cap - Sludge and debris in the bottom few feet of the basin will be encapsulated between the basin floor and overlying fill material to isolate if from the environment; (3) Demolishing the above grade portion of the structure and relocating the resulting debris to another location or disposing of the debris in-place; and (4) Capping the basin area with a concrete slab which is part of an engineered cap to prevent inadvertent intrusion. The estimated total grout volume to fill the 105-R Reactor Disassembly Basin is 24,424 cubic meters or 31,945 cubic yards. Portland cement-based structural fill materials were design and tested for the reactor ISD project and a placement strategy for stabilizing the basin was developed. Based on structural engineering analyses and work flow considerations, the recommended maximum lift height is 5 feet with 24 hours between lifts. Pertinent data and information related to the SRS 105-R-Reactor Disassembly Basin in-situ decommissioning include: regulatory documentation, residual water management, area preparation activities, technology needs, fill material designs and testing, and fill placement strategy. This information is applicable to decommissioning both the 105-P and 105-R facilities. The ISD process for the entire 105-P and 105-R reactor facilities will require approximately 250,000 cubic yards (191,140 cubic meters) of grout and 2,400 cubic yards (1,840 cubic meters) of structural concrete which will be placed over a twelve month period to meet the accelerated schedule ISD schedule. The status and lessons learned in the SRS Reactor Facility ISD process will be described.

  15. Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1994-12-31T23:59:59.000Z

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future.

  16. Modular Pebble Bed Reactor High Temperature Gas Reactor

    E-Print Network [OSTI]

    Built · Site Assembled · On--line Refueling · Modules added to meet demand. · No Reprocessing · High Experiments · Non-Proliferation · Safeguards · Waste Disposal · Reactor Research/ Demonstration Facility

  17. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-06-01T23:59:59.000Z

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full-core-height diodes. Moreover, placing the fuel on the outside of the diode makes possible reactors with much higher fuel volume fractions, which enable power-flattened fast reactors scalable to very low power levels without the need for life-limiting hydride moderators or the use of efficiency-limiting driver fuel. In addition, with the fuel on the outside its swelling does not increase the emitter diameter or reduce the interelectrode gap. This should permit long lifetimes even with closer spacings, which can significantly improve the system efficiences. This was confirmed by coupled neutronic, thermal, thermionic, and electrical system analyses - some of which are presented in this paper - and by subsequent experiments. A companion paper presented next describes the fabrication and testing of full-scale converter elements, both fueled and unfueled, and summarizes the test results obtained. There is a duplicate copy in the file.

  18. Benchmark for evaluation and validation of reactor simulations (BEAVRS)

    SciTech Connect (OSTI)

    Horelik, N.; Herman, B.; Forget, B.; Smith, K. [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)

    2013-07-01T23:59:59.000Z

    Advances in parallel computing have made possible the development of high-fidelity tools for the design and analysis of nuclear reactor cores, and such tools require extensive verification and validation. This paper introduces BEAVRS, a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading patterns, and numerous in-vessel components. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from fifty-eight instrumented assemblies. Initial comparisons between calculations performed with MIT's OpenMC Monte Carlo neutron transport code and measured cycle 1 HZP test data are presented, and these results display an average deviation of approximately 100 pcm for the various critical configurations and control rod worth measurements. Computed HZP radial fission detector flux maps also agree reasonably well with the available measured data. All results indicate that this benchmark will be extremely useful in validation of coupled-physics codes and uncertainty quantification of in-core physics computational predictions. The detailed BEAVRS specification and its associated data package is hosted online at the MIT Computational Reactor Physics Group web site (http://crpg.mit.edu/), where future revisions and refinements to the benchmark specification will be made publicly available. (authors)

  19. Discovery of Isotopes of Elements with Z $\\ge$ 100

    E-Print Network [OSTI]

    M. Thoennessen

    2012-03-09T23:59:59.000Z

    Currently, 163 isotopes of elements with Z $\\ge$ 100 have been observed and the discovery of these isotopes is discussed here. For each isotope a brief synopsis of the first refereed publication, including the production and identification method, is presented.

  20. Gamma Spectrum from Neutron Capture on Tungsten Isotopes

    E-Print Network [OSTI]

    Hurst, Aaron

    2011-01-01T23:59:59.000Z

    FROM NEUTRON CAPTURE ON TUNGSTEN ISOTOPES A. M. HURST ?1,2 ,capture on the stable tungsten isotopes is presented, withknown decay schemes of the tungsten isotopes from neutron

  1. Designing Reactors to Facilitate Decommissioning

    SciTech Connect (OSTI)

    Richard H. Meservey

    2006-06-01T23:59:59.000Z

    Critics of nuclear power often cite issues with tail-end-of-the-fuel-cycle activities as reasons to oppose the building of new reactors. In fact, waste disposal and the decommissioning of large nuclear reactors have proven more challenging than anticipated. In the early days of the nuclear power industry the design and operation of various reactor systems was given a great deal of attention. Little effort, however, was expended on end-of-the-cycle activities, such as decommissioning and disposal of wastes. As early power and test reactors have been decommissioned difficulties with end-of-the-fuel-cycle activities have become evident. Even the small test reactors common at the INEEL were not designed to facilitate their eventual decontamination, decommissioning, and dismantlement. The results are that decommissioning of these facilities is expensive, time consuming, relatively hazardous, and generates large volumes of waste. This situation clearly supports critics concerns about building a new generation of power reactors.

  2. Fast quench reactor and method

    DOE Patents [OSTI]

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.

    1998-05-12T23:59:59.000Z

    A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

  3. Tropical Pacific nutrient dynamics in the modern and pleistocene ocean : insights from the nitrogen isotope system

    E-Print Network [OSTI]

    Rafter, Patrick Anthony

    2009-01-01T23:59:59.000Z

    nitrogen isotopes, an enrichment that is conventionallyisotopes upon denitrification also imparts a strong isotopic enrichment

  4. Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover; David A. Petti

    2008-10-01T23:59:59.000Z

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The design of the first experiment (designated AGR-1) was completed in 2005, and the fabrication and assembly of the test train as well as the support systems and fission product monitoring system that monitor and control the experiment during irradiation were completed in September 2006. The experiment was inserted in the ATR in December 2006, and is serving as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed and the status of the experiment is provided.

  5. Nuclear reactor with internal thimble-type delayed neutron detection system

    SciTech Connect (OSTI)

    Gross, Kenny C. (Lemont, IL); Poloncsik, John (Downers Grove, IL); Lambert, John D. B. (Wheaton, IL)

    1990-01-01T23:59:59.000Z

    This invention teaches improved apparatus for the method of detecting a breach in cladded fuel used in a nuclear reactor. The detector apparatus is located in the primary heat exchanger which conveys part of the reactor coolant past at least three separate delayed-neutron detectors mounted in this heat exchanger. The detectors are spaced apart such that the coolant flow time from the core to each detector is different, and these differences are known. The delayed-neutron activity at the detectors is a function of the delay time after the reaction in the fuel until the coolant carrying the delayed-neutron emitter passes the respective detector. This time delay is broken down into separate components including an isotopic holdup time required for the emitter to move through the fuel from the reaction to the coolant at the breach, and two transit times required for the emitter now in the coolant to flow from the breach to the detector loop and then via the loop to the detector. At least two of these time components are determined during calibrated operation of the reactor. Thereafter during normal reactor operation, repeated comparisons are made by the method of regression approximation of the third time component for the best-fit line correlating measured delayed-neutron activity against activity that is approximated according to specific equations. The equations use these time-delay components and known parameter values of the fuel and of the part and emitting daughter isotopes.

  6. Neutrino Oscillation Studies with Reactors

    E-Print Network [OSTI]

    Petr Vogel; Liangjian Wen; Chao Zhang

    2015-03-03T23:59:59.000Z

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  7. Neutrino Oscillation Studies with Reactors

    E-Print Network [OSTI]

    Petr Vogel; Liangjian Wen; Chao Zhang

    2015-04-27T23:59:59.000Z

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  8. Neutrino Oscillation Studies with Reactors

    E-Print Network [OSTI]

    Vogel, Petr; Zhang, Chao

    2015-01-01T23:59:59.000Z

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  9. Progress Update: Reactor Disassembly Grouting

    ScienceCinema (OSTI)

    Cody, Tom

    2012-06-14T23:59:59.000Z

    Grouting the P&R reactors in order to remove these basins as an environmental threat. This will end the Cold War legacy and end the environmental footprint.

  10. Progress Update: Reactor Disassembly Grouting

    SciTech Connect (OSTI)

    Cody, Tom

    2010-01-01T23:59:59.000Z

    Grouting the P&R reactors in order to remove these basins as an environmental threat. This will end the Cold War legacy and end the environmental footprint.

  11. Neutrino oscillation studies with reactors

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

    Vogel, P. [California Inst. of Technology (CalTech), Pasadena, CA (United States). Kellog Radiation Lab.; Wen, L.J. [Chinese Academy of Sciences (CAS), Beijing (China). Inst. of High Energy Physics (IHEP); Zhang, C. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-04-27T23:59:59.000Z

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavours are quantum mechanical mixtures. Over the past several decades, reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle ?13. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  12. Thermonuclear Reflect AB-Reactor

    E-Print Network [OSTI]

    Alexander Bolonkin

    2008-03-26T23:59:59.000Z

    The author offers a new kind of thermonuclear reflect reactor. The remarkable feature of this new reactor is a three net AB reflector, which confines the high temperature plasma. The plasma loses part of its energy when it contacts with the net but this loss can be compensated by an additional permanent plasma heating. When the plasma is rarefied (has a small density), the heat flow to the AB reflector is not large and the temperature in the triple reflector net is lower than 2000 - 3000 K. This offered AB-reactor has significantly less power then the currently contemplated power reactors with magnetic or inertial confinement (hundreds-thousands of kW, not millions of kW). But it is enough for many vehicles and ships and particularly valuable for tunnelers, subs and space apparatus, where air to burn chemical fuel is at a premium or simply not available. The author has made a number of innovations in this reactor, researched its theory, developed methods of computation, made a sample computation of typical project. The main point of preference for the offered reactor is its likely cheapness as a power source. Key words: Micro-thermonuclear reactor, Multi-reflex AB-thermonuclear reactor, Self-magnetic AB-thermonuclear reactor, aerospace thermonuclear engine.

  13. Spatial periphery of lithium isotopes

    SciTech Connect (OSTI)

    Galanina, L. I., E-mail: galan_lidiya@mail.ru; Zelenskaja, N. S. [Moscow State University, Skobeltsyn Institute of Nuclear Physics (Russian Federation)

    2013-12-15T23:59:59.000Z

    The spatial structure of lithium isotopes is studied with the aid of the charge-exchange and (t, p) reactions on lithium nuclei. It is shown that an excited isobaric-analog state of {sup 6}Li (0{sup +}, 3.56MeV) has a halo structure formed by a proton and a neutron, that, in the {sup 9}Li nucleus, there is virtually no neutron halo, and that {sup 11}Li is a Borromean nucleus formed by a {sup 9}Li core and a two-neutron halo manifesting itself in cigar-like and dineutron configurations.

  14. Isotopic Scaling in Nuclear Reactions

    E-Print Network [OSTI]

    M. B. Tsang; W. A. Friedman; C. K. Gelbke; W. G. Lynch; G. Verde; H. Xu

    2001-03-26T23:59:59.000Z

    A three parameter scaling relationship between isotopic distributions for elements with Z$\\leq 8$ has been observed that allows a simple description of the dependence of such distributions on the overall isospin of the system. This scaling law (termed iso-scaling) applies for a variety of reaction mechanisms that are dominated by phase space, including evaporation, multifragmentation and deeply inelastic scattering. The origins of this scaling behavior for the various reaction mechanisms are explained. For multifragmentation processes, the systematics is influenced by the density dependence of the asymmetry term of the equation of state.

  15. Stable carbon and nitrogen isotope enrichment in primate tissues

    E-Print Network [OSTI]

    Crowley, Brooke E.; Carter, Melinda L.; Karpanty, Sarah M.; Zihlman, Adrienne L.; Koch, Paul L.; Dominy, Nathaniel J.

    2010-01-01T23:59:59.000Z

    carbon and nitrogen isotope enrichment in primate tissuesfactor (a) and isotope enrichment values (e), which provideisotope values from different modern primate tissues. Additionally, using these mean apparent enrichment

  16. aluminium isotopes: Topics by E-print Network

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

    CERN Preprints Summary: Currently, 31 actinium, 31 thorium, 28 protactinium, and 23 uranium isotopes have so far been observed; the discovery of these isotopes is discussed. For...

  17. americium isotopes: Topics by E-print Network

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

    CERN Preprints Summary: Currently, 31 actinium, 31 thorium, 28 protactinium, and 23 uranium isotopes have so far been observed; the discovery of these isotopes is discussed. For...

  18. activated bismuth isotopes: Topics by E-print Network

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

    CERN Preprints Summary: Currently, 31 actinium, 31 thorium, 28 protactinium, and 23 uranium isotopes have so far been observed; the discovery of these isotopes is discussed. For...

  19. astatine isotopes: Topics by E-print Network

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

    CERN Preprints Summary: Currently, 31 actinium, 31 thorium, 28 protactinium, and 23 uranium isotopes have so far been observed; the discovery of these isotopes is discussed. For...

  20. A Helium Isotope Perspective On The Dixie Valley, Nevada, Hydrothermal...

    Open Energy Info (EERE)

    Helium Isotope Perspective On The Dixie Valley, Nevada, Hydrothermal System Jump to: navigation, search OpenEI Reference LibraryAdd to library Journal Article: A Helium Isotope...

  1. Isotopic Analysis- Fluid At Dixie Valley Geothermal Area (Kennedy...

    Open Energy Info (EERE)

    geothermal resources with deep, fault hosted permeable fluid flow pathways and the helium Isotopic composition of the surface fluids. The authors suggest that helium isotopes...

  2. Fractionation of Oxygen Isotopes in Phosphate during its Interactions...

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

    Fractionation of Oxygen Isotopes in Phosphate during its Interactions with Iron Oxides. Fractionation of Oxygen Isotopes in Phosphate during its Interactions with Iron Oxides....

  3. applied isotope techniques: Topics by E-print Network

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

    M. 22 Applied Radiation and Isotopes 55 (2001) 707713 Bronchial dosimeter for radon progeny Biology and Medicine Websites Summary: Applied Radiation and Isotopes 55 (2001)...

  4. Freese-casting as a Novel Manufacturing Process for Fast Reactor Fuels

    SciTech Connect (OSTI)

    Wegst, Ulrike G.K.; Allen, Todd; Sridharan, Kumar

    2014-04-07T23:59:59.000Z

    Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reqctors requires novel fuel types based on new materials and designs that can acieve higher performance requirements (higher burn up, higher power, and greator margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a welldefined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

  5. Reactor coolant pump flywheel

    DOE Patents [OSTI]

    Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

    2013-11-26T23:59:59.000Z

    A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

  6. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, J.E.; Meuschke, R.E.

    1995-05-02T23:59:59.000Z

    A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

  7. Nuclear reactor control assembly

    SciTech Connect (OSTI)

    Negron, S.B.

    1991-06-11T23:59:59.000Z

    This patent describes an assembly for providing global power control in a nuclear reactor having the core split into two halves. It comprises a disk assembly formed from at least two disks each machined with an identical surface hole pattern such that rotation of one disk relative to the other causes the hole pattern to open or close, the disk assembly being positioned substantially at the longitudinal center of and coaxial with the core halves; and means for rotating at least one of the disks relative to the other.

  8. Nuclear reactor control apparatus

    SciTech Connect (OSTI)

    Sridhar, B.N.

    1981-08-28T23:59:59.000Z

    Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additonal magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

  9. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.; Marasco, J.A.

    1996-02-27T23:59:59.000Z

    A fluidized bed reactor system is described which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves. 3 figs.

  10. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN)

    1993-01-01T23:59:59.000Z

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves.

  11. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

    1995-01-01T23:59:59.000Z

    A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves.

  12. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

    1996-01-01T23:59:59.000Z

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves.

  13. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.

    1993-12-14T23:59:59.000Z

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase is described. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figures.

  14. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, James E. (Greensburg, PA); Meuschke, Robert E. (Pittsburgh, PA)

    1995-01-01T23:59:59.000Z

    A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

  15. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.; Marasco, J.A.

    1995-04-25T23:59:59.000Z

    A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figs.

  16. Nuclear reactor control apparatus

    DOE Patents [OSTI]

    Sridhar, Bettadapur N. (Cupertino, CA)

    1983-11-01T23:59:59.000Z

    Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additional magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

  17. N Reactor - Hanford Site

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated Codes |IsLove Your1 SECTIONES2008-54174MoreMuseum Day atMyNERSCVanN Reactor

  18. Advanced Mass Spectrometers for Hydrogen Isotope Analyses

    SciTech Connect (OSTI)

    Chastagner, P.

    2001-08-01T23:59:59.000Z

    This report is a summary of the results of a joint Savannah River Laboratory (SRL) - Savannah River Plant (SRP) ''Hydrogen Isotope Mass Spectrometer Evaluation Program''. The program was undertaken to evaluate two prototype hydrogen isotope mass spectrometers and obtain sufficient data to permit SRP personnel to specify the mass spectrometers to replace obsolete instruments.

  19. Positive and inverse isotope effect on superconductivity

    E-Print Network [OSTI]

    Tian De Cao

    2009-09-04T23:59:59.000Z

    This article improves the BCS theory to include the inverse isotope effect on superconductivity. An affective model can be deduced from the model including electron-phonon interactions, and the phonon-induced attraction is simply and clearly explained on the electron Green function. The focus of this work is on how the positive or inverse isotope effect occurs in superconductors.

  20. MARK E. CONRAD Center for Isotope Geochemistry

    E-Print Network [OSTI]

    Ajo-Franklin, Jonathan

    with stable isotopes, fate and transport of groundwater contaminants, vadose zone hydrology, metal uptake by lichens, paleoclimatic patterns in California and stable isotope systematics of clay minerals. Geologist of gold property in the Mother Lode of California. Exploration Geologist (Anaconda Minerals Company; 6

  1. The Quest for the Heaviest Uranium Isotope

    E-Print Network [OSTI]

    S. Schramm; D. Gridnev; D. V. Tarasov; V. N. Tarasov; W. Greiner

    2012-01-17T23:59:59.000Z

    We study Uranium isotopes and surrounding elements at very large neutron number excess. Relativistic mean field and Skyrme-type approaches with different parametrizations are used in the study. Most models show clear indications for isotopes that are stable with respect to neutron emission far beyond N=184 up to the range of around N=258.

  2. Isotope separation by selective photodissociation of glyoxal

    DOE Patents [OSTI]

    Marling, John B. (Pleasanton, CA)

    1976-01-01T23:59:59.000Z

    Dissociation products, mainly formaldehyde and carbon monoxide, enriched in a desired isotope of carbon, oxygen, or hydrogen are obtained by the selective photodissociation of glyoxal wherein glyoxal is subjected to electromagnetic radiation of a predetermined wavelength such that photon absorption excites and induces dissociation of only those molecules of glyoxal containing the desired isotope.

  3. [Carbon isotope fractionation inplants]. Final report

    SciTech Connect (OSTI)

    O`Leary, M.H.

    1990-12-31T23:59:59.000Z

    The objectives of this research are: To develop a theoretical and experimental framework for understanding isotope fractionations in plants; and to develop methods for using this isotope fractionation for understanding the dynamics of CO{sub 2} fixation in plants. Progress is described.

  4. 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)

    2013-07-01T23:59:59.000Z

    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)

  5. Plutonium and minor actinide utilisation in a pebble-bed high temperature reactor

    SciTech Connect (OSTI)

    Petrov, B. Y.; Kuijper, J. C.; Oppe, J.; De Haas, J. B. M. [Nuclear Research and Consultancy Group, Westerduinweg 3, 1755 ZG Petten (Netherlands)

    2012-07-01T23:59:59.000Z

    This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA reactor loaded with plutonium and minor actinide (Pu/MA) fuel. Starting from knowledge and experience gained in the Euratom FP5 projects HTR-N and HTR-N1, this study aims at demonstrating the potential of high temperature reactors to utilize or transmute Pu/MA fuel. The work has been performed within the Euratom FP6 project PUMA. A number of different fuel types and fuel configurations have been analyzed and compared with respect to incineration performance and safety-related reactor parameters. The results show the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burnup of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burnup and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the 'wallpaper' fuel does not have advantage over the standard fuel design in this respect. After 100 years of decay following the fuel discharge, the total amount of actinides remains almost unchanged for all of the fuel types considered. Among the plutonium isotopes, only the amount of Pu-241 is reduced significantly due to its relatively short half-life. (authors)

  6. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect (OSTI)

    R. Wigeland; J. Cahalan

    2009-09-01T23:59:59.000Z

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and thermophysical properties of the fuel and their compatibility with the reactor coolant, with corresponding differences in the challenges presented to the reactor developers. Accident phenomena are discussed for the sodium-cooled fast reactor based on the mechanistic progression of conditions from accident initiation to accident termination, whether a benign state is achieved or more severe consequences are expected. General principles connecting accident phenomena and fuel properties are developed from the oxide and metal fuel safety analyses, providing guidelines that can be used as part of the evaluation for selection of fuel type for the sodium-cooled fast reactor.

  7. Activities of ?-ray emitting isotopes in rainwater from Greater Sudbury, Canada following the Fukushima incident

    E-Print Network [OSTI]

    B. T. Cleveland; F. A. Duncan; I. T. Lawson; N. J. T. Smith; E. Vazquez-Jauregui

    2012-02-29T23:59:59.000Z

    We report the activity measured in rainwater samples collected in the Greater Sudbury area of eastern Canada on 3, 16, 20, and 26 April 2011. The samples were gamma-ray counted in a germanium detector and the isotopes 131I and 137Cs, produced by the fission of 235U, and 134Cs, produced by neutron capture on 133Cs, were observed at elevated levels compared to a reference sample of ice-water. These elevated activities are ascribed to the accident at the Fukushima Dai-ichi nuclear reactor complex in Japan that followed the 11 March earthquake and tsunami. The activity levels observed at no time presented health concerns.

  8. Granular Dynamics in Pebble Bed Reactor Cores

    E-Print Network [OSTI]

    Laufer, Michael Robert

    2013-01-01T23:59:59.000Z

    in a pebble-bed nuclear reactor,” Phys. Rev. E, vol. 74, no.cycles of the pebble bed reactor,” Nuclear Engineering andoptimization of pebble-bed reactors,” Annals of Nuclear

  9. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    SciTech Connect (OSTI)

    NONE

    1994-10-01T23:59:59.000Z

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition.

  10. Gaseous reactor control system

    SciTech Connect (OSTI)

    Abdel-Khalik, S.

    1991-09-03T23:59:59.000Z

    This paper describes a nuclear reactor control system for controlling the reactivity of the core of a nuclear reactor. It includes a control gas having a high neutron cross-section; a first tank containing a first supply of the control gas; a first conduit providing a first fluid passage extending into the core, the first conduit being operatively connected to communicate with the first tank; a first valve operatively connected to regulate the flow of the control gas between the first tank and the first conduit; a second conduit concentrically disposed around the first conduit such that a second fluid passage is defined between the outer surface of the first conduit and the inner surface of the second conduit; a second tank containing a second supply of the control gas, the second tank being operatively connected to communicate with the second fluid passage; a second supply valve operatively connected to regulate the flow of the control gas between the second tank and the second fluid passage.

  11. Solvent refined coal reactor quench system

    DOE Patents [OSTI]

    Thorogood, R.M.

    1983-11-08T23:59:59.000Z

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

  12. ISOTOPES

    E-Print Network [OSTI]

    Lederer, C. Michael

    2013-01-01T23:59:59.000Z

    constructed to enrich liquid UF6 slightly as feed for thej) b. Optimum a. s: .X. UF6 feed, (kg per year) XBL 7912 -

  13. ISOTOPES

    E-Print Network [OSTI]

    Lederer, C. Michael

    2013-01-01T23:59:59.000Z

    lithium hydroxide (56, 14), A plant utilizing this reaction to produce 1000 kg per year of 99.99% 7Li at a price

  14. ISOTOPES

    E-Print Network [OSTI]

    Lederer, C. Michael

    2013-01-01T23:59:59.000Z

    1978). United States Gas Centrifuge Program for 72. Energy~~;h)radius in a gas centrifuge with va 400 m/s. P 0 and U68,14) of a gas centrifuge with radius = 0.09145 m, length=

  15. ISOTOPES

    E-Print Network [OSTI]

    Lederer, C. Michael

    2013-01-01T23:59:59.000Z

    the columns used to separate UF5, which had a separativeof uranium by dissociation of UF5, by multiple vibrational

  16. Reactor Neutrino Physics -- An Update

    E-Print Network [OSTI]

    Felix Boehm

    1999-06-18T23:59:59.000Z

    We review the status and the results of reactor neutrino experiments. Long baseline oscillation experiments at Palo Verde and Chooz have provided limits for the oscillation parameters while the recently proposed Kamland experiment at a baseline of more than 100km is now in the planning stage. We also describe the status of neutrino magnetic moment experiments at reactors.

  17. Estimates of Radioxenon Released from Southern Hemisphere Medical isotope Production Facilities Using Measured Air Concentrations and Atmospheric Transport Modeling

    SciTech Connect (OSTI)

    Eslinger, Paul W.; Friese, Judah I.; Lowrey, Justin D.; McIntyre, Justin I.; Miley, Harry S.; Schrom, Brian T.

    2014-04-06T23:59:59.000Z

    Abstract The International Monitoring System (IMS) of the Comprehensive-Nuclear-Test-Ban-Treaty monitors the atmosphere for radioactive xenon leaking from underground nuclear explosions. Emissions from medical isotope production represent a challenging background signal when determining whether measured radioxenon in the atmosphere is associated with a nuclear explosion prohibited by the treaty. The Australian Nuclear Science and Technology Organisation (ANSTO) operates a reactor and medical isotope production facility in Lucas Heights, Australia. This study uses two years of release data from the ANSTO medical isotope production facility and Xe-133 data from three IMS sampling locations to estimate the annual releases of Xe-133 from medical isotope production facilities in Argentina, South Africa, and Indonesia. Atmospheric dilution factors derived from a global atmospheric transport model were used in an optimization scheme to estimate annual release values by facility. The annual releases of about 6.8×1014 Bq from the ANSTO medical isotope production facility are in good agreement with the sampled concentrations at these three IMS sampling locations. Annual release estimates for the facility in South Africa vary from 1.2×1016 to 2.5×1016 Bq and estimates for the facility in Indonesia vary from 6.1×1013 to 3.6×1014 Bq. Although some releases from the facility in Argentina may reach these IMS sampling locations, the solution to the objective function is insensitive to the magnitude of those releases.

  18. Atomic vapor laser isotope separation of lead-210 isotope

    DOE Patents [OSTI]

    Scheibner, K.F.; Haynam, C.A.; Johnson, M.A.; Worden, E.F.

    1999-08-31T23:59:59.000Z

    An isotopically selective laser process and apparatus for removal of Pb-210 from natural lead that involves a one-photon near-resonant, two-photon resonant excitation of one or more Rydberg levels, followed by field ionization and then electrostatic extraction. The wavelength to the near-resonant intermediate state is counter propagated with respect to the second wavelength required to populate the final Rydberg state. This scheme takes advantage of the large first excited state cross section, and only modest laser fluences are required. The non-resonant process helps to avoid two problems: first, stimulated Raman Gain due to the nearby F=3/2 hyperfine component of Pb-207 and, second, direct absorption of the first transition process light by Pb-207. 5 figs.

  19. Atomic vapor laser isotope separation of lead-210 isotope

    DOE Patents [OSTI]

    Scheibner, Karl F. (Tracy, CA); Haynam, Christopher A. (Pleasanton, CA); Johnson, Michael A. (Pleasanton, CA); Worden, Earl F. (Diablo, CA)

    1999-01-01T23:59:59.000Z

    An isotopically selective laser process and apparatus for removal of Pb-210 from natural lead that involves a one-photon near-resonant, two-photon resonant excitation of one or more Rydberg levels, followed by field ionization and then electrostatic extraction. The wavelength to the near-resonant intermediate state is counter propagated with respect to the second wavelength required to populate the final Rydberg state. This scheme takes advantage of the large first excited state cross section, and only modest laser fluences are required. The non-resonant process helps to avoid two problems: first, stimulated Raman Gain due to the nearby F=3/2 hyperfine component of Pb-207 and, second, direct absorption of the first transition process light by Pb-207.

  20. Radiation Shielding for Fusion Reactors

    SciTech Connect (OSTI)

    Santoro, R.T.

    1999-10-01T23:59:59.000Z

    Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel.

  1. Granular Dynamics in Pebble Bed Reactor Cores

    E-Print Network [OSTI]

    Laufer, Michael Robert

    2013-01-01T23:59:59.000Z

    gas reactors with the effective heat transfer of a molten salt coolant and the passive natural circulation safety systems of sodium fast reactors.

  2. MOOSE simulating nuclear reactor CRUD buildup

    ScienceCinema (OSTI)

    None

    2014-07-21T23:59:59.000Z

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  3. MOOSE simulating nuclear reactor CRUD buildup

    SciTech Connect (OSTI)

    None

    2014-02-06T23:59:59.000Z

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  4. The Endurance Bioenergy Reactor | Argonne National Laboratory

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

    The Endurance Bioenergy Reactor Share Description Argonne biophysicist Dr. Philip Laible and Air Force Major Matt Michaud talks about he endurance bioenergy reactor-a device that...

  5. Fast reactors and nuclear nonproliferation

    SciTech Connect (OSTI)

    Avrorin, E.N. [Russian Federal Nuclear Center - Zababakhin Institute of Applied Physics, Snezhinsk (Russian Federation); Rachkov, V.I.; Chebeskov, A.N. [State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering, Bondarenko Square, 1, Obninsk, Kaluga region, 249033 (Russian Federation)

    2013-07-01T23:59:59.000Z

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

  6. Automatic reactor power control for a pressurized water reactor

    SciTech Connect (OSTI)

    Jungin Choi (Kyungwon Univ. (Korea, Republic of)); Yungjoon Hah (Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)); Unchul Lee (Seoul National Univ. (Korea, Republic of))

    1993-05-01T23:59:59.000Z

    An automatic reactor power control system is presented for a pressurized water reactor (PWR). The associated reactor control strategy is called mode K.' The new system implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial shape change, which allows automatic control of the axial power distribution. Thus, the mode K enables precise regulation of both the reactivity and the power distribution, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load-follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1,000-MW (electric) PWR. The simulation results illustrate that the mode K would be a practical reactor power control strategy for the increased automation of nuclear plants.

  7. Evaluation of Homogeneous Options: Effects of Minor Actinide Exclusion from Single and Double Tier Recycle in Sodium Fast Reactors

    SciTech Connect (OSTI)

    R. M. Ferrer; S. Bays; M. Pope

    2008-03-01T23:59:59.000Z

    The Systems Analysis Campaign under the Global Nuclear Energy Partnership (GNEP) has requested the fuel cycle analysis group at the Idaho National Laboratory (INL) to analyze and provide isotopic data for four scenarios in which different strategies for Minor Actinides (MA) management are investigated. A 1000 MWth commercial-scale Sodium Fast Reactor (SFR) design was selected as the baseline in this scenario study. Two transuranic (TRU) conversion ratios, defined as the ratio of the amount of TRU produced over the TRU destroyed in the reactor core, along with different fuel-types were investigated.

  8. Study of a multi-beam accelerator driven thorium reactor

    SciTech Connect (OSTI)

    Ludewig, H.; Aronson, A.

    2011-03-01T23:59:59.000Z

    The primary advantages that accelerator driven systems have over critical reactors are: (1) Greater flexibility regarding the composition and placement of fissile, fertile, or fission product waste within the blanket surrounding the target, and (2) Potentially enhanced safety brought about by operating at a sufficiently low value of the multiplication factor to preclude reactivity induced events. The control of the power production can be achieved by vary the accelerator beam current. Furthermore, once the beam is shut off the system shuts down. The primary difference between the operation of an accelerator driven system and a critical system is the issue of beam interruptions of the accelerator. These beam interruptions impose thermo-mechanical loads on the fuel and mechanical components not found in critical systems. Studies have been performed to estimate an acceptable number of trips, and the value is significantly less stringent than had been previously estimated. The number of acceptable beam interruptions is a function of the length of the interruption and the mission of the system. Thus, for demonstration type systems and interruption durations of 1sec < t < 5mins, and t > 5mins 2500/yr and 50/yr are deemed acceptable. However, for industrial scale power generation without energy storage type systems and interruption durations of t < 1sec., 1sec < t < 10secs., 10secs < t < 5mins, and t > 5mins, the acceptable number of interruptions are 25000, 2500, 250, and 3 respectively. However, it has also been concluded that further development is required to reduce the number of trips. It is with this in mind that the following study was undertaken. The primary focus of this study will be the merit of a multi-beam target system, which allows for multiple spallation sources within the target/blanket assembly. In this manner it is possible to ameliorate the effects of sudden accelerator beam interruption on the surrounding reactor, since the remaining beams will still be supplying source neutrons. The proton beam will be assumed to have an energy of 1 GeV, and the target material will be natural lead, which will also be the coolant for the reactor assembly. Three proton beam arrangements will be considered, first a single beam (the traditional arrangement) with an entry at the assembly center, two more options will consist of three and six entry locations. The reactor fuel assembly parameters will be based on those of the S-PRISM fast reactor proposed by GE, and the fuel composition and type will be based on that proposed by Aker Solutions for use in their accelerator driven thorium reactor. The following table summarizes the parameters to be used in this study. The isotopic composition of the fertile material is 100% Th-232, and the plutonium isotopic distribution corresponds to that characteristic of the discharge from a typical LWR, following five years of decay. Thus, the isotopic distribution for the plutonium is; Pu-238 2.5%, Pu-239 53.3%, Pu-240 25.1%, Pu-241 11.8%, and Pu-242 7.3%.

  9. Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1994-12-31T23:59:59.000Z

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials.

  10. University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor

    SciTech Connect (OSTI)

    Eric C. Woolstenhulme; Dana M. Hewit

    2008-09-01T23:59:59.000Z

    The Department of Energy’s Idaho National Laboratory, under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at Purdue University Reactor. With this work completed and in anticipation of other impending conversion projects, the INL convened and engaged the project participants in a structured discussion to capture the lessons learned. The lessons learned process has allowed us to capture gaps, opportunities, and good practices, drawing from the project team’s experiences. These lessons will be used to raise the standard of excellence, effectiveness, and efficiency in all future conversion projects.

  11. Nuclear reactor control

    DOE Patents [OSTI]

    Cawley, William E. (Phoenix, AZ); Warnick, Robert F. (Pasco, WA)

    1982-01-01T23:59:59.000Z

    1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

  12. Nuclear reactor control apparatus

    DOE Patents [OSTI]

    Sridhar, Bettadapur N. (Cupertino, CA)

    1983-10-25T23:59:59.000Z

    Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the safety rod to release the rod from the drive shaft. In preferred embodiments of the invention, an additional latch is provided to release a lower portion of a safety rod under conditions that may interfere with movement of the entire rod.

  13. Nuclear reactor control

    SciTech Connect (OSTI)

    Cawley, W.E.; Warnick, R.F.

    1982-03-30T23:59:59.000Z

    In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

  14. Isotopic composition of Silurian seawater

    SciTech Connect (OSTI)

    Knauth, L.P.; Kealy, S.; Larimer, S.

    1985-01-01T23:59:59.000Z

    Direct isotopic analyses of 21 samples of the Silurian hydrosphere preserved as fluid inclusions in Silurian halite deposits in the Michigan Basin Salina Group yield delta/sup 18/O, deltaD ranging from 0.2 to +5.9 and -26 to -73, respectively. delta/sup 18/O has the same range as observed for modern halite facies evaporite waters and is a few per thousand higher than 100 analyses of fluid inclusions in Permian halite. deltaD is about 20 to 30 per thousand lower than modern and Permian examples. The trajectory of evaporating seawater on a deltaD-delta/sup 18/O diagram initially has a positive slope of 3-6, but hooks strongly downward to negative values, the shape of the hook depending upon humidity. Halite begins to precipitate at delta values similar to those observed for the most /sup 18/O rich fluid inclusions. Subsequent evaporation yields progressively more negative delta values as observed for the fluid inclusions. The fluid inclusion data can be readily explained in terms of evaporating seawater and are consistent with the degree of evaporation deduced from measured bromide profiles. These data are strongly inconsistent with arguments that Silurian seawater was 5.5 per thousand depleted in /sup 18/O. delta/sup 18/O for evaporite waters is systematically related to that of seawater, and does not show a -5.5 per thousand shift in the Silurian, even allowing for variables which affect the isotope evaporation trajectory. The lower deltaD may indicate a component of gypsum dehydration waters or may suggest a D-depleted Silurian hydrosphere.

  15. Evolutionary/advanced light water reactor data report

    SciTech Connect (OSTI)

    NONE

    1996-02-09T23:59:59.000Z

    The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (``burned``) in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ``evolutionary`` or ``advanced`` designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ``evolutionary`` LWR alternative.

  16. Comparison of isotopic transmutation modelling codes

    E-Print Network [OSTI]

    Beard, Carl Allen

    1990-01-01T23:59:59.000Z

    . Numerical solutions to the equations that govern isotopic transmutation can be obtained in several ways. Each method possesses a certain amount of intrinsic error which is inherent in the solution scheme, but which can also vary depending upon... of the removed nuclide if it falls between two long-lived isotopes, or by adding the initial concentration of the short-lived isotope to the first long-lived nuclide which occurs in the production chain. In this second case, the final contribution from...

  17. Theory of the helium isotope shift

    E-Print Network [OSTI]

    Yerokhin, Vladimir A

    2015-01-01T23:59:59.000Z

    Theory of the isotope shift of the centroid energies of light few-electron atoms is reviewed. Numerical results are presented for the isotope shift of the $2^3P$-$2^3S$ and $2^1S$-$2^3S$ transition energies of $^3$He and $^4$He. By comparing theoretical predictions for the isotope shift with the experimental results, the difference of the squares of the nuclear charge radii of $^3$He and $^4$He, $\\delta R^2$, is determined with high accuracy.

  18. Helium isotopes in historical lavas from Mount Vesuvius Comment on `Noble gas isotopic ratios from historical lavas

    E-Print Network [OSTI]

    Graham, David W.

    Discussion Helium isotopes in historical lavas from Mount Vesuvius Comment on `Noble gas isotopic. Introduction Helium isotope results recently published by Tedesco et al. [1] appear to show a decrease in 3 He. Results Helium isotope results from our laboratory are reported in Table 1. The 3 He/4 He ratio has been

  19. Preliminary Notice of Violation, International Isotopes Idaho...

    Broader source: Energy.gov (indexed) [DOE]

    to Work Planning and Control Deficiencies associated with Replacement of Exhaust Ventilation Filters at the Test Reactor Area Hot Cell Facility at the Idaho National...

  20. United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support

    SciTech Connect (OSTI)

    Douglas Morrell

    2011-03-01T23:59:59.000Z

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  1. Reactor protection system design alternatives for sodium fast reactors

    E-Print Network [OSTI]

    DeWitte, Jacob D. (Jacob Dominic)

    2011-01-01T23:59:59.000Z

    Historically, unprotected transients have been viewed as design basis events that can significantly challenge sodium-cooled fast reactors. The perceived potential consequences of a severe unprotected transient in a ...

  2. Reactor physics design of supercritical CO?-cooled fast reactors

    E-Print Network [OSTI]

    Pope, Michael A. (Michael Alexander)

    2004-01-01T23:59:59.000Z

    Gas-Cooled Fast Reactors (GFRs) are among the GEN-IV designs proposed for future deployment. Driven by anticipated plant cost reduction, the use of supercritical CO? (S-CO?) as a Brayton cycle working fluid in a direct ...

  3. Nuclear reactor downcomer flow deflector

    DOE Patents [OSTI]

    Gilmore, Charles B. (Greensburg, PA); Altman, David A. (Pittsburgh, PA); Singleton, Norman R. (Murrysville, PA)

    2011-02-15T23:59:59.000Z

    A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

  4. FORIG: a computer code for calculating radionuclide generation and depletion in fusion and fission reactors. User's manual

    SciTech Connect (OSTI)

    Blink, J.A.

    1985-03-01T23:59:59.000Z

    In this manual we describe the use of the FORIG computer code to solve isotope-generation and depletion problems in fusion and fission reactors. FORIG runs on a Cray-1 computer and accepts more extensive activation cross sections than ORIGEN2 from which it was adapted. This report is an updated and a combined version of the previous ORIGEN2 and FORIG manuals. 7 refs., 15 figs., 13 tabs.

  5. Diffusion Coefficient of Tritium Through Molten Salt Flibe and Rate of Tritium Leak from Fusion Reactor System

    SciTech Connect (OSTI)

    Fukada, Satoshi [Kyushu University (Japan); Anderl, Robert A. [Idaho National Engineering and Environmental Laboratory (United States); Sagara, Akio [National Institute for Fusion Science (Japan); Nishikawa, Masabumi [Kyushu University (Japan)

    2005-07-15T23:59:59.000Z

    Diffusion coefficients of hydrogen isotopes in Flibe were correlated with making reference to previous relating data of F{sup -} ion self-diffusivity and Flibe viscosity and so on. Rates of tritium permeation through structural materials in a fusion reactor system with Flibe blanket were estimated comparatively under conditions with or without a Flibe permeation barrier. A way to lower the tritium leak rate below a level regulated by law was proposed, and its effectiveness was discussed.

  6. 2012 Annual Report Research Reactor Infrastructure Program

    SciTech Connect (OSTI)

    Douglas Morrell

    2012-11-01T23:59:59.000Z

    The content of this report is the 2012 Annual Report for the Research Reactor Infrastructure Program.

  7. Advanced Test Reactor National Scientific User Facility Partnerships

    SciTech Connect (OSTI)

    Frances M. Marshall; Todd R. Allen; Jeff B. Benson; James I. Cole; Mary Catherine Thelen

    2012-03-01T23:59:59.000Z

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin-Madison; (8) Illinois Institute of Technology (IIT) Materials Research Collaborative Access Team (MRCAT) beamline at Argonne National Laboratory's Advanced Photon Source; and (9) Nanoindenter in the University of California at Berkeley (UCB) Nuclear Engineering laboratory Materials have been analyzed for ATR NSUF users at the Advanced Photon Source at the MRCAT beam, the NIST Center for Neutron Research in Gaithersburg, MD, the Los Alamos Neutron Science Center, and the SHaRE user facility at Oak Ridge National Laboratory (ORNL). Additionally, ORNL has been accepted as a partner facility to enable ATR NSUF users to access the facilities at the High Flux Isotope Reactor and related facilities.

  8. Degradation of Isotopic Lactate and Acetate

    E-Print Network [OSTI]

    Aronoff, S.

    2010-01-01T23:59:59.000Z

    prescribed, BaC03 from the degradation of Ba acetnte co~~above procedure by the degradation of sjnthetic radio-lacticNo. W-7405-Eng o -48 DEGRADATION OF ISOTOPIC LACTATE AND

  9. The Isotopic Abundances of Magnesium in Stars

    E-Print Network [OSTI]

    Pamela Gay; David L. Lambert

    1999-11-11T23:59:59.000Z

    Isotopic abundance ratios 24^Mg:25^Mg:26^Mg are derived for 20 stars from high- resolution spectra of the MgH A-X 0-0 band at 5140AA. With the exception of the weak g-band giant HR 1299, the stars are dwarfs that sample the metallicity range -1.8 < [Fe/H] <0.0. The abundance of 25^Mg amd 26^Mg relative to the dominant isotope 24^Mg decreases with decreasing [Fe/H] in fair accord with predictions from a recent model of galactic chemical evolution in which the Mg isotopes are synthesised by massive stars. Several stars appear especially enriched in the heavier Mg isotopes suggesting contamination by material from the envelopes of intermediate-mass AGB stars.

  10. Nonequilibrium clumped isotope signals in microbial methane

    E-Print Network [OSTI]

    Wang, David T.

    Methane is a key component in the global carbon cycle with a wide range of anthropogenic and natural sources. Although isotopic compositions of methane have traditionally aided source identification, the abundance of its ...

  11. A SUPERCONDUCTING-SOLENOID ISOTOPE SPECTROMETER

    E-Print Network [OSTI]

    O'Donnell, Tom

    A SUPERCONDUCTING-SOLENOID ISOTOPE SPECTROMETER FOR PRODUCTION OF NEUTRON-RICH NUCLEI ( 136 Xe Superconducting Cyclotron Laboratory's weekly \\Green Sheet," 30 July 1999 #12; c Thomas W. O'Donnell 2000 All

  12. How far is a Fusion Power Reactor from an Experimental Reactor?

    E-Print Network [OSTI]

    1 How far is a Fusion Power Reactor from an Experimental Reactor? R. Toschi(1) , P. Barabaschi(2 September 2000 Madrid, Spain #12;2 How far is a fusion power reactor from an experimental reactor? R. Toschi To support a request of very substantial resources to build and operate an experimental reactor such as ITER

  13. Carbon isotope fractionation in autotrophic Chromatium

    E-Print Network [OSTI]

    Wong, William Wai-Lun

    1974-01-01T23:59:59.000Z

    CARSON ISOTOPE FRACTIONATION IN AUTOTPOPHIC CHROYATIUN A Thesis 'JILLIAJJ J JAI LJJN BONG Submitted to the Graduate College of Texas A&H University in partial fulfillment of the requirenent for the degree of PLASTER OF SCIENCE August 1974...) August 1974 ABSTRACT Carbon Isotope Fractionation in Autotrophic Chromatium (August 1974) blilliam Wai-Lun Wang, B. S. , Texas Lutheran College Co-Chairmen of Advisory Committee: Dr. Isilliam N. Sackett Dr. Chauncey P. . Benedict Bacterial cells...

  14. Mass Parameterizations and Predictions of Isotopic Observables

    E-Print Network [OSTI]

    S. R. Souza; P. Danielewicz; S. Das Gupta; R. Donangelo; W. A. Friedman; W. G. Lynch; W. P. Tan; M. B. Tsang

    2003-03-24T23:59:59.000Z

    We discuss the accuracy of mass models for extrapolating to very asymmetric nuclei and the impact of such extrapolations on the predictions of isotopic observables in multifragmentation. We obtain improved mass predictions by incorporating measured masses and extrapolating to unmeasured masses with a mass formula that includes surface symmetry and Coulomb terms. We find that using accurate masses has a significant impact on the predicted isotopic observables.

  15. Terracentric Nuclear Fission Reactor: Background, Basis, Feasibility, Structure, Evidence, and Geophysical Implications

    E-Print Network [OSTI]

    J. Marvin Herndon

    2013-12-31T23:59:59.000Z

    The background, basis, feasibility, structure, evidence, and geophysical implications of a naturally occurring Terracentric nuclear fission georeactor are reviewed. For a nuclear fission reactor to exist at the center of the Earth, all of the following conditions must be met: (1) There must originally have been a substantial quantity of uranium within Earth's core; (2) There must be a natural mechanism for concentrating the uranium; (3) The isotopic composition of the uranium at the onset of fission must be appropriate to sustain a nuclear fission chain reaction; (4) The reactor must be able to breed a sufficient quantity of fissile nuclides to permit operation over the lifetime of Earth to the present; (5) There must be a natural mechanism for the removal of fission products; (6) There must be a natural mechanism for removing heat from the reactor; (7) There must be a natural mechanism to regulate reactor power level, and; (8) The location of the reactor or must be such as to provide containment and prevent meltdown. Herndon's georeactor alone is shown to meet those conditions. Georeactor existence evidence based upon helium measurements and upon antineutrino measurements is described. Geophysical implications discussed include georeactor origin of the geomagnetic field, geomagnetic reversals from intense solar outbursts and severe Earth trauma, as well as georeactor heat contributions to global dynamics.

  16. SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configuration

    SciTech Connect (OSTI)

    Gauld, I.C.

    2000-03-16T23:59:59.000Z

    Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the United States, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k{sub eff} values within about 1% {Delta}k/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models.

  17. Terracentric Nuclear Fission Reactor: Background, Basis, Feasibility, Structure, Evidence, and Geophysical Implications

    E-Print Network [OSTI]

    Herndon, J Marvin

    2013-01-01T23:59:59.000Z

    The background, basis, feasibility, structure, evidence, and geophysical implications of a naturally occurring Terracentric nuclear fission georeactor are reviewed. For a nuclear fission reactor to exist at the center of the Earth, all of the following conditions must be met: (1) There must originally have been a substantial quantity of uranium within Earth's core; (2) There must be a natural mechanism for concentrating the uranium; (3) The isotopic composition of the uranium at the onset of fission must be appropriate to sustain a nuclear fission chain reaction; (4) The reactor must be able to breed a sufficient quantity of fissile nuclides to permit operation over the lifetime of Earth to the present; (5) There must be a natural mechanism for the removal of fission products; (6) There must be a natural mechanism for removing heat from the reactor; (7) There must be a natural mechanism to regulate reactor power level, and; (8) The location of the reactor or must be such as to provide containment and prevent ...

  18. Distribution ICategory: General Reactor Technology

    E-Print Network [OSTI]

    Shlyakhter, Ilya

    --- Distribution ICategory: General Reactor Technology (UC-520) ANl-92/23 AR(;ONNE NATIONAL, progressively by Huygens, Maxwell and Roentgen, mankind has learned to observe it, measure it, control it

  19. Spectral shift reactor control method

    SciTech Connect (OSTI)

    Impink, A.J. Jr.

    1987-08-18T23:59:59.000Z

    The method is described of closely controlling the reactor water coolant temperature of an operating spectral-shift nuclear reactor, the reactor comprising a core formed of fuel assemblies through which the reactor water coolant flows; different types of elongated elements operable to be controllably moved into and out of the core; one type of the elongated elements comprising control rods formed of neutron absorbing material and operable to decrease reactivity through neutron absorption when inserted into the core; another of the types of elongated elements comprising displacer rods formed of material which has a low absorption for neutrons and which have overall neutron-absorbing and moderating characteristics essentially not exceeding those of hollow tubular Zircaloy members with a filling zirconium oxide or aluminum oxide, the displacer rods operating to displace an equivalent volume of water coolant fluid from the core when inserted therein to decrease reactivity and to increase reactivity when moved from the core.

  20. University Reactor Matching Grants Program

    SciTech Connect (OSTI)

    John Valentine; Farzad Rahnema; Said Abdel-Khalik

    2003-02-14T23:59:59.000Z

    During the 2002 Fiscal year, funds from the DOE matching grant program, along with matching funds from the industrial sponsors, have been used to support research in the area of thermal-hydraulics. Both experimental and numerical research projects have been performed. Experimental research focused on two areas: (1) Identification of the root cause mechanism for axial offset anomaly in pressurized water reactors under prototypical reactor conditions, and (2) Fluid dynamic aspects of thin liquid film protection schemes for inertial fusion reactor chambers. Numerical research focused on two areas: (1) Multi-fluid modeling of both two-phase and two-component flows for steam conditioning and mist cooling applications, and (2) Modeling of bounded Rayleigh-Taylor instability with interfacial mass transfer and fluid injection through a porous wall simulating the ''wetted wall'' protection scheme in inertial fusion reactor chambers. Details of activities in these areas are given.