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Sample records for flux isotope reactor

  1. OSTIblog Articles in the High Flux Isotope Reactor Topic | OSTI...

    Office of Scientific and Technical Information (OSTI)

    High Flux Isotope Reactor Topic The NXS Class of 2014 by Kathy Chambers 19 Nov, 2014 in ... National Laboratory, High Flux Isotope Reactor, National School on Neutron and X-ray ...

  2. Revision of HFIR (High Flux Isotope Reactor) operating procedures

    SciTech Connect (OSTI)

    McGinty, D.M.

    1987-01-23

    This report documents modifications to the facility and changes in some operating procedures for the High Flux Isotope Reactor (HFIR). The topics covered include: Reactor Operation, Reactor Start-up, Reactor Safety Systems, Reactor Control Systems, Reporting Requirements, and Administrative Procedures. (FI)

  3. 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.

  4. 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.

  5. 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.

  6. 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.

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

    SciTech Connect (OSTI)

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

    2012-07-01

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

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

    SciTech Connect (OSTI)

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

    2012-01-01

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

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

    Office of Science (SC) Website

    Facilities » High Flux Isotope Reactor (HFIR) Scientific User Facilities (SUF) Division SUF Home About User Facilities X-Ray Light Sources Neutron Scattering Facilities Spallation Neutron Source (SNS) High Flux Isotope Reactor (HFIR) Nanoscale Science Research Centers (NSRCs) Projects Accelerator & Detector Research Science Highlights Principal Investigators' Meetings BES Home Neutron Scattering Facilities High Flux Isotope Reactor (HFIR) Print Text Size: A A A FeedbackShare Page Quick

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

    SciTech Connect (OSTI)

    Sease, J.D.

    1998-03-01

    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.

  11. High Flux Isotope Reactor system RELAP5 input model

    SciTech Connect (OSTI)

    Morris, D.G.; Wendel, M.W.

    1993-01-01

    A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model.

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

    SciTech Connect (OSTI)

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

    1998-05-01

    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.

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

    SciTech Connect (OSTI)

    Banfield, J.; Mervin, B.; Hart, S.; Ritchie, J.; Walker, S.; Ruggles, A.; Maldonado, G. I.

    2012-07-01

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

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

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

    Office of Energy Efficiency and Renewable Energy (EERE)

    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.

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

    Office of Energy Efficiency and Renewable Energy (EERE)

    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.

  6. 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.

  7. 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.

  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, 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.

  10. 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.

  11. 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.

  12. High Flux Isotope Reactor | Neutron Science at ORNL

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

    HFIR is also used for medical, industrial, and research isotope production; research on severe neutron damage to materials; and neutron activation analysis to examine trace ...

  13. Homogeneous fast-flux isotope-production reactor

    DOE Patents [OSTI]

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

    1982-08-19

    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.

  14. The ORNL High Flux Isotope Reactor and New Advanced Fuel Testing Capabilities

    SciTech Connect (OSTI)

    Ott, Larry J; McDuffee, Joel Lee

    2011-01-01

    The U.S. Department of Energy s High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), was originally designed (in the 1960s) primarily as a part of the overall program to produce transuranic isotopes for use in the heavy-element research program of the United States. Today, the reactor is a highly versatile machine, producing medical and transuranic isotopes and performing materials test experimental irradiations and neutron-scattering experiments. The ability to test advanced fuels and cladding materials in a thermal neutron spectrum in the United States is limited, and a fast-spectrum irradiation facility does not currently exist in this country. The HFIR has a distinct advantage for consideration as a fuel/cladding irradiation facility because of the extremely high neutron fluxes that this reactor provides over the full thermal- to fast-neutron energy range. New test capabilities have been developed that will allow testing of advanced nuclear fuels and cladding materials in the HFIR under prototypic light-water reactor (LWR) and fast-reactor (FR) operating conditions.

  15. OSTIblog Articles in the High Flux Isotope Reactor Topic | OSTI, US Dept of

    Office of Scientific and Technical Information (OSTI)

    Energy Office of Scientific and Technical Information High Flux Isotope Reactor Topic The NXS Class of 2014 by Kathy Chambers 19 Nov, 2014 in Every summer for the past 16 years, the Department of Energy has invited the best and brightest graduates from across the country to attend the National School on Neutron and X-ray Scattering (NXS). This year, 65 graduate students attending North American universities, and studying physics, chemistry, materials science, or related fields, participated

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

    SciTech Connect (OSTI)

    Jain, Prashant K; Freels, James D

    2015-01-01

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

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

    SciTech Connect (OSTI)

    Cook, David Howard

    2009-01-01

    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.

  18. Optimization of Depletion Modeling and Simulation for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Betzler, Benjamin R; Ade, Brian J; Chandler, David; Ilas, Germina; Sunny, Eva E

    2015-01-01

    Monte Carlo based depletion tools used for the high-fidelity modeling and simulation of the High Flux Isotope Reactor (HFIR) come at a great computational cost; finding sufficient approximations is necessary to make the use of these tools feasible. The optimization of the neutronics and depletion model for the HFIR is based on two factors: (i) the explicit representation of the involute fuel plates with sets of polyhedra and (ii) the treatment of depletion mixtures and control element position during depletion calculations. A very fine representation (i.e., more polyhedra in the involute plate approximation) does not significantly improve simulation accuracy. The recommended representation closely represents the physical plates and ensures sufficient fidelity in regions with high flux gradients. Including the fissile targets in the central flux trap of the reactor as depletion mixtures has the greatest effect on the calculated cycle length, while localized effects (e.g., the burnup of specific isotopes or the power distribution evolution over the cycle) are more noticeable consequences of including a critical control element search or depleting burnable absorbers outside the fuel region.

  19. Neutron scattering at the high flux isotope reactor at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Yethiraj, M.; Fernandez-Baca, J.A.

    1995-03-01

    Since its beginnings in Oak Ridge and Argonne in the late 1940`s, neutron scattering has been established as the premier tool to study matter in its various states. Since the thermal neutron wavelength is of the same order of magnitude as typical atomic spacings and because they have comparable energies to those of atomic excitations in solids, both structure and dynamics of matter can be studied via neutron scattering. The High Flux Isotope Reactor (HFIR) provides an intense source of neutrons with which to carry out these measurements. This paper summarizes the available neutron scattering facilities at the HFIR.

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

    SciTech Connect (OSTI)

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

    2010-10-01

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

  1. Development of a Scale Model for High Flux Isotope Reactor Cycle 400

    SciTech Connect (OSTI)

    Ilas, Dan

    2012-03-01

    The development of a comprehensive SCALE computational model for the High Flux Isotope Reactor (HFIR) is documented and discussed in this report. The SCALE model has equivalent features and functionality as the reference MCNP model for Cycle 400 that has been used extensively for HFIR safety analyses and for HFIR experiment design and analyses. Numerical comparisons of the SCALE and MCNP models for the multiplication constant, power density distribution in the fuel, and neutron fluxes at several locations in HFIR indicate excellent agreement between the results predicted with the two models. The SCALE HFIR model is presented in sufficient detail to provide the users of the model with a tool that can be easily customized for various safety analysis or experiment design requirements.

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

    SciTech Connect (OSTI)

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

    2009-01-01

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

  3. High Flux Isotope Reactor Core Analysis-Challenges and Recent Enhancements in Modeling and Simulation

    SciTech Connect (OSTI)

    Ilas, Germina

    2016-01-01

    A concerted effort over the past few years has focused on enhancing the core depletion models for the High Flux Isotope Reactor (HFIR) as part of a comprehensive study for designing a HFIR core that would use low-enriched uranium (LEU) fuel. A HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed for use as a reference for the design of an LEU fuel for HFIR and to improve the basis for analyses that support HFIR s current operation with high-enriched uranium (HEU) fuel. This paper summarizes the recent improvements in modeling and simulation for HFIR core analyses, with a focus on core depletion models.

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

    SciTech Connect (OSTI)

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

    2014-09-01

    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.

  5. A system analysis computer model for the High Flux Isotope Reactor (HFIRSYS Version 1)

    SciTech Connect (OSTI)

    Sozer, M.C.

    1992-04-01

    A system transient analysis computer model (HFIRSYS) has been developed for analysis of small break loss of coolant accidents (LOCA) and operational transients. The computer model is based on the Advanced Continuous Simulation Language (ACSL) that produces the FORTRAN code automatically and that provides integration routines such as the Gear`s stiff algorithm as well as enabling users with numerous practical tools for generating Eigen values, and providing debug outputs and graphics capabilities, etc. The HFIRSYS computer code is structured in the form of the Modular Modeling System (MMS) code. Component modules from MMS and in-house developed modules were both used to configure HFIRSYS. A description of the High Flux Isotope Reactor, theoretical bases for the modeled components of the system, and the verification and validation efforts are reported. The computer model performs satisfactorily including cases in which effects of structural elasticity on the system pressure is significant; however, its capabilities are limited to single phase flow. Because of the modular structure, the new component models from the Modular Modeling System can easily be added to HFIRSYS for analyzing their effects on system`s behavior. The computer model is a versatile tool for studying various system transients. The intent of this report is not to be a users manual, but to provide theoretical bases and basic information about the computer model and the reactor.

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

    SciTech Connect (OSTI)

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

    2009-12-01

    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.

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

    SciTech Connect (OSTI)

    Lastres, Oscar; Chandler, David; Jarrell, Joshua J; Maldonado, G. Ivan

    2011-01-01

    The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is a versatile 85 MW{sub th}, pressurized, light water-cooled and -moderated research reactor. The core consists of two fuel elements, an inner fuel element (IFE) and an outer fuel element (OFE), each constructed of involute fuel plates containing high-enriched-uranium (HEU) fuel ({approx}93 wt% {sup 235}U/U) in the form of U{sub 3}O{sub 8} in an Al matrix and encapsulated in Al-6061 clad. An over-moderated flux trap is located in the center of the core, a large beryllium reflector is located on the outside of the core, and two control elements (CE) are located between the fuel and the reflector. The flux trap and reflector house numerous experimental facilities which are used for isotope production, material irradiation, and cold/thermal neutron scattering. Over the past five decades, the US Department of Energy (DOE) and its agencies have been producing radioisotope power systems used by the National Aeronautics and Space Administration (NASA) for unmanned, long-term space exploration missions. Plutonium-238 is used to power Radioisotope Thermoelectric Generators (RTG) because it has a very long half-life (t{sub 1/2} {approx} 89 yr.) and it generates about 0.5 watts/gram when it decays via alpha emission. Due to the recent shortage and uncertainty of future production, the DOE has proposed a plan to the US Congress to produce {sup 238}Pu by irradiating {sup 237}Np as early as in fiscal year 2011. An annual production rate of 1.5 to 2.0 kg of {sup 238}Pu is expected to satisfy these needs and could be produced in existing national nuclear facilities like HFIR and the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Reactors at the Savannah River Site were used in the past for {sup 238}Pu production but were shut down after the last production in 1988. The nation's {sup 237}Np inventory is currently stored at INL. A plan for producing {sup 238}Pu at US research reactor

  8. 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

    2013-11-01

    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 -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.

  9. Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling

    SciTech Connect (OSTI)

    Travis, Adam R

    2014-05-01

    A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.

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

    SciTech Connect (OSTI)

    Ilas, Germina; Chandler, David; Ade, Brian J; Sunny, Eva E; Betzler, Benjamin R; Pinkston, Daniel

    2015-03-01

    A concerted effort over the past few years has been focused on enhancing the core model for the High Flux Isotope Reactor (HFIR), as part of a comprehensive study for HFIR conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel. At this time, the core model used to perform analyses in support of HFIR operation is an MCNP model for the beginning of Cycle 400, which was documented in detail in a 2005 technical report. A HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed to serve as reference for the design of an LEU fuel for HFIR. The recent enhancements in modeling and simulations for HFIR that are discussed in the present report include: (1) revision of the 2005 MCNP model for the beginning of Cycle 400 to improve the modeling data and assumptions as necessary based on appropriate primary reference sources HFIR drawings and reports; (2) improvement of the fuel region model, including an explicit representation for the involute fuel plate geometry that is characteristic to HFIR fuel; and (3) revision of the Monte Carlo-based depletion model for HFIR in use since 2009 but never documented in detail, with the development of a new depletion model for the HFIR explicit fuel plate representation. The new HFIR models for Cycle 400 are used to determine various metrics of relevance to reactor performance and safety assessments. The calculated metrics are compared, where possible, with measurement data from preconstruction critical experiments at HFIR, data included in the current HFIR safety analysis report, and/or data from previous calculations performed with different methods or codes. The results of the analyses show that the models presented in this report provide a robust and reliable basis for HFIR analyses.

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

    SciTech Connect (OSTI)

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

    2015-01-01

    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

  12. Reactor Physics Studies of Reduced-Tantaulum-Content Control and Safety Elements for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Primm, R.T., III

    2003-11-01

    Some of the unirradiated High Flux Isotope Reactor (HFIR) control elements discharged during the late 1990s were observed to have cladding damage--local swelling or blistering. The cladding damage was limited to the tantalum/europium interface of the element and is thought to result from interaction of hydrogen and europium to form a compound of lower density than europium oxide, thus leading to a ''blistering'' of the control plate cladding. Reducing the tantalum loading in the control plates should help preclude this phenomena. The impact of the change to the control plates on the operation of the reactor was assessed. Regarding nominal, steady-state reactor operation, the impact of the change in the power distribution in the core due to reduced tantalum content was calculated and found to be insignificant. The magnitude and impact of the change in differential control element worth was calculated, and the differential worths of reduced tantalum elements vs the current elements from equivalent-burnup critical configurations were determined to be unchanged within the accuracy of the computational method and relevant experimental measurements. The location of the critical control elements symmetric positions for reduced tantalum elements was found to be 1/3 in. less withdrawn relative to existing control elements regardless of the value of fuel cycle burnup (time in the fuel cycle). The magnitude and impact of the change in the shutdown margin (integral rod worth) was assessed and found to be unchanged. Differential safety element worth values for the reduced-tantalum-content elements were calculated for postulated accident conditions and were found to be greater than values currently assumed in HFIR safety analyses.

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

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01

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

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

    SciTech Connect (OSTI)

    Ilas, Germina; Primm, Trent

    2009-11-01

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

  15. Delivery of completed irradiation vehicles and the quality assurance document to the High Flux Isotope Reactor for irradiation

    SciTech Connect (OSTI)

    Petrie, Christian M.; McDuffee, Joel Lee; Katoh, Yutai; Terrani, Kurt A.

    2015-10-01

    This report details the initial fabrication and delivery of two Fuel Cycle Research and Development (FCRD) irradiation capsules (ATFSC01 and ATFSC02), with associated quality assurance documentation, to the High Flux Isotope Reactor (HFIR). The capsules and documentation were delivered by September 30, 2015, thus meeting the deadline for milestone M3FT-15OR0202268. These irradiation experiments are testing silicon carbide composite tubes in order to obtain experimental validation of thermo-mechanical models of stress states in SiC cladding irradiated under a prototypic high heat flux. This document contains a copy of the completed capsule fabrication request sheets, which detail all constituent components, pertinent drawings, etc., along with a detailed summary of the capsule assembly process performed by the Thermal Hydraulics and Irradiation Engineering Group (THIEG) in the Reactor and Nuclear Systems Division (RNSD). A complete fabrication package record is maintained by the THIEG and is available upon request.

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

    SciTech Connect (OSTI)

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

    2010-02-01

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

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

    SciTech Connect (OSTI)

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

    2009-03-01

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

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

    SciTech Connect (OSTI)

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

    2011-02-01

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

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

    SciTech Connect (OSTI)

    Primm, Trent; Guida, Tracey

    2010-02-01

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

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

    SciTech Connect (OSTI)

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

    2012-03-01

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

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

    SciTech Connect (OSTI)

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

    2007-11-01

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

  2. 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

    2010-01-01

    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.

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

    SciTech Connect (OSTI)

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

    2006-11-01

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

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

    SciTech Connect (OSTI)

    Daily, Charles R.

    2015-10-01

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

  5. High flux reactor

    DOE Patents [OSTI]

    Lake, James A.; Heath, Russell L.; Liebenthal, John L.; DeBoisblanc, Deslonde R.; Leyse, Carl F.; Parsons, Kent; Ryskamp, John M.; Wadkins, Robert P.; Harker, Yale D.; Fillmore, Gary N.; Oh, Chang H.

    1988-01-01

    A high flux reactor is comprised of a core which is divided into two symetric segments housed in a pressure vessel. The core segments include at least one radial fuel plate. The spacing between the plates functions as a coolant flow channel. The core segments are spaced axially apart such that a coolant mixing plenum is formed between them. A channel is provided such that a portion of the coolant bypasses the first core section and goes directly into the mixing plenum. The outlet coolant from the first core segment is mixed with the bypass coolant resulting in a lower inlet temperature to the lower core segment.

  6. Design and Nuclear-Safety Related Simulations of Bare-Pellet Test Irradiations for the Production of Pu-238 in the High Flux Isotope Reactor using COMSOL

    SciTech Connect (OSTI)

    Freels, James D; Jain, Prashant K; Hobbs, Randy W

    2012-01-01

    The Oak Ridge National Laboratory (ORNL)is developing technology to produce plutonium-238 for the National Aeronautics and Space Administration (NASA) as a power source material for powering vehicles while in deep-space[1]. The High Flux Isotope Reactor (HFIR) of ORNL has been utilized to perform test irradiations of incapsulated neptunium oxide (NpO2) and aluminum powder bare pellets for purposes of understanding the performance of the pellets during irradiation[2]. Post irradiation examinations (PIE) are currently underway to assess the effect of temperature, thermal expansion, swelling due to gas production, fission products, and other phenomena

  7. Production of Medical Radioisotopes in the ORNL High Flux Isotope Reactor (HFIR) for Cancer Treatment and Arterial Restenosis Therapy after PTCA

    DOE R&D Accomplishments [OSTI]

    Knapp, F. F. Jr.; Beets, A. L.; Mirzadeh, S.; Alexander, C. W.; Hobbs, R. L.

    1998-06-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) represents an important resource for the production of a wide variety of medical radioisotopes. In addition to serving as a key production site for californium-252 and other transuranic elements, important examples of therapeutic radioisotopes which are currently routinely produced in the HFIR for distribution include dysprosium-166 (parent of holmium-166), rhenium-186, tin-117m and tungsten-188 (parent of rhenium-188). The nine hydraulic tube (HT) positions in the central high flux region permit the insertion and removal of targets at any time during the operating cycle and have traditionally represented a major site for production of medical radioisotopes. To increase the irradiation capabilities of the HFIR, special target holders have recently been designed and fabricated which will be installed in the six Peripheral Target Positions (PTP), which are also located in the high flux region. These positions are only accessible during reactor refueling and will be used for long-term irradiations, such as required for the production of tin-117m and tungsten-188. Each of the PTP tubes will be capable of housing a maximum of eight HT targets, thus increasing the total maximum number of HT targets from the current nine, to a total of 57. In this paper the therapeutic use of reactor-produced radioisotopes for bone pain palliation and vascular brachytherapy and the therapeutic medical radioisotope production capabilities of the ORNL HFIR are briefly discussed.

  8. Experimental Plan and Irradiation Target Design for FeCrAl Embrittlement Screening Tests Conducted Using the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Field, Kevin G.; Howard, Richard H.; Yamamoto, Yukinori

    2015-06-26

    The objective of the FeCrAl embrittlement screening tests being conducted through the use of Oak Ridge National Laboratories (ORNL) High Flux Isotope Reactor is to provide data on the radiation-induced changes in the mechanical properties including radiation-induced hardening and embrittlement through systematic testing and analysis. Data developed on the mechanical properties will be supported by extensive microstructural evaluations to assist in the development of structure-property relationships and provide a sound, fundamental understanding of the performance of FeCrAl alloys in intense neutron radiation fields. Data and analysis developed as part of this effort will be used to assist in the determination of FeCrAl alloys as a viable material for commercial light water reactor (LWR) applications with a primary focus as an accident tolerant cladding.

  9. Neutronics Simulations of 237Np Targets to Support Safety-Basis and 238Pu Production Assessment Efforts at the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Chandler, David; Ellis, Ronald James

    2015-01-01

    Fueled by two highly enriched uranium-bearing fuel elements surrounded by a large concentric ring of beryllium reflector, the High Flux Isotope Reactor (HFIR) provides one of the highest neutron fluxes in the world and is used to produce unique isotopes like plutonium-238. The National Aeronautics and Space Administration use radioisotope thermoelectric generators powered by 238Pu for deep-space missions. As part of the US Department of Energy s task to reestablish the domestic production of 238Pu, a technology demonstration sub-project has been initiated to establish a new 238Pu supply chain. HFIR safety-basis neutronics calculations are being performed to ensure the target irradiations have no adverse impacts on reactor performance and to calculate data required as input to follow-on thermal-structural, thermal-hydraulic and radionuclide/dose analyses. Plutonium-238 production assessments are being performed to estimate the amount of 238Pu that can be produced in HFIR s permanent beryllium reflector. It is estimated that a total of 0.96 1.12 kg 238Pu (~1.28 1.49 kg PuO2 at 85% 238Pu/Pu purity) could be produced per year in HFIR s permanent beryllium reflector irradiation facilities if they are all utilized.

  10. Study of the Potential Impact of Gamma-Induced Radiolytic Gases on Loading of Cesium Onto Crystalline Silicotitanate Sorbent at ORNL's High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Mattus, A.J.

    2001-02-12

    The use of an engineered form of crystalline silicotitanate as a potential sorbent for the removal and concentration of cesium from the high-level waste at the Savannah River Site was investigated. Results conclusively showed this sorbent to be unaffected by gamma-induced radiolytic gas formation during column loading. Closely controlled column-loading experiments were performed at the Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR) in a gamma field with a conservative dose rate expected to exceed that in a full-scale column by a factor of nearly 16. Operation of column loading under expected nominal full-scale field conditions in the HFIR pool showed that radiolytic gases were formed at a previously calculated generation rate of 0.4 mL per liter of feed solution. When the resulting cesium-loading curve in the gamma field was compared with that of a control experiment in the absence of a gamma field, no discernable difference in the curves (within analytical error) was detected. Both curves were in good agreement with the VERSE computer-generated curve. Results conclusively indicate that the production of radiolytic gases within a full-scale column is not expected to result in reduced capacity or associated gas generation problems during operation at the Savannah River Site.

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

    SciTech Connect (OSTI)

    Not Available

    1994-03-01

    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.

  12. Sandia National Laboratories Medical Isotope Reactor concept.

    SciTech Connect (OSTI)

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

    2010-04-01

    This report describes the Sandia National Laboratories Medical Isotope Reactor and hot cell facility concepts. The reactor proposed is designed to be capable of producing 100% of the U.S. demand for the medical isotope {sup 99}Mo. The concept is novel in that the fuel for the reactor and the targets for the {sup 99}Mo production are the same. There is no driver core required. The fuel pins that are in the reactor core are processed on a 7 to 21 day irradiation cycle. The fuel is low enriched uranium oxide enriched to less than 20% {sup 235}U. The fuel pins are approximately 1 cm in diameter and 30 to 40 cm in height, clad with Zircaloy (zirconium alloy). Approximately 90 to 150 fuel pins are arranged in the core in a water pool {approx}30 ft deep. The reactor power level is 1 to 2 MW. The reactor concept is a simple design that is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days. The fuel fabrication, reactor design and operation, and {sup 99}Mo production processing use well-developed technologies that minimize the technological and licensing risks. There are no impediments that prevent this type of reactor, along with its collocated hot cell facility, from being designed, fabricated, and licensed today.

  13. Small-Scale Reactor for the Production of Medical Isotopes -...

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

    Small-Scale Reactor for the Production of Medical Isotopes Sandia National Laboratories ... Out LEU reactor is ready to construct -US government is looking for investors. We have ...

  14. HFBR handbook, 1992: High flux beam reactor

    SciTech Connect (OSTI)

    Axe, J.D.; Greenberg, R.

    1992-10-01

    Welcome to the High Flux Beam Reactor (HFBR), one of the world premier neutron research facilities. This manual is intended primarily to acquaint outside users (and new Brookhaven staff members) with (almost) everything they need to know to work at the HFBR and to help make the stay at Brookhaven pleasant as well as profitable. Safety Training Programs to comply with US Department of Energy (DOE) mandates are in progress at BNL. There are several safety training requirements which must be met before users can obtain unescorted access to the HFBR. The Reactor Division has prepared specific safety training manuals which are to be sent to experimenters well in advance of their expected arrival at BNL to conduct experiments. Please familiarize yourself with this material and carefully pay strict attention to all the safety and security procedures that are in force at the HFBR. Not only your safety, but the continued operation of the facility, depends upon compliance.

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

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

    Research and Medical ... Research and Medical Isotope Reactor Supply Our goal is to fuel research and test reactors with low-enriched uranium. Y-12 tops the short list of the ...

  16. Reactor Fuel Isotopics and Code Validation for Nuclear Applications

    SciTech Connect (OSTI)

    Francis, Matthew W.; Weber, Charles F.; Pigni, Marco T.; Gauld, Ian C.

    2015-02-01

    Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for 133Cs and 153Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied

  17. Spheromak reactor with poloidal flux-amplifying transformer

    DOE Patents [OSTI]

    Furth, Harold P.; Janos, Alan C.; Uyama, Tadao; Yamada, Masaaki

    1987-01-01

    An inductive transformer in the form of a solenoidal coils aligned along the major axis of a flux core induces poloidal flux along the flux core's axis. The current in the solenoidal coil is then reversed resulting in a poloidal flux swing and the conversion of a portion of the poloidal flux to a toroidal flux in generating a spheromak plasma wherein equilibrium approaches a force-free, minimum Taylor state during plasma formation, independent of the initial conditions or details of the formation. The spheromak plasma is sustained with the Taylor state maintained by oscillating the currents in the poloidal and toroidal field coils within the plasma-forming flux core. The poloidal flux transformer may be used either as an amplifier stage in a moving plasma reactor scenario for initial production of a spheromak plasma or as a method for sustaining a stationary plasma and further heating it. The solenoidal coil embodiment of the poloidal flux transformer can alternately be used in combination with a center conductive cylinder aligned along the length and outside of the solenoidal coil. This poloidal flux-amplifying inductive transformer approach allows for a relaxation of demanding current carrying requirements on the spheromak reactor's flux core, reduces plasma contamination arising from high voltage electrode discharge, and improves the efficiency of poloidal flux injection.

  18. Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

    SciTech Connect (OSTI)

    Cartier, J.; Casoli, P.; Chappert, F.

    2013-07-01

    In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

  19. Dry phase reactor for generating medical isotopes

    DOE Patents [OSTI]

    Mackie, Thomas Rockwell; Heltemes, Thad Alexander

    2016-05-03

    An apparatus for generating medical isotopes provides for the irradiation of dry-phase, granular uranium compounds which are then dissolved in a solvent for separation of the medical isotope from the irradiated compound. Once the medical isotope is removed, the dissolved compound may be reconstituted in dry granular form for repeated irradiation.

  20. Neutron flux profile monitor for use in a fission reactor

    DOE Patents [OSTI]

    Kopp, Manfred K.; Valentine, Kenneth H.

    1983-01-01

    A neutron flux monitor is provided which consists of a plurality of fission counters arranged as spaced-apart point detectors along a delay line. As a fission event occurs in any one of the counters, two delayed current pulses are generated at the output of the delay line. The time separation of the pulses identifies the counter in which the particular fission event occured. Neutron flux profiles of reactor cores can be more accurately measured as a result.

  1. Scoping assessment on medical isotope production at the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Scott, S.W.

    1997-08-29

    The Scoping Assessment addresses the need for medical isotope production and the capability of the Fast Flux Test Facility to provide such isotopes. Included in the discussion are types of isotopes used in radiopharmaceuticals, which types of cancers are targets, and in what way isotopes provide treatment and/or pain relief for patients.

  2. RADIATION DOSIMETRY AT THE BNL HIGH FLUX BEAM REACTOR AND MEDICAL RESEARCH REACTOR.

    SciTech Connect (OSTI)

    HOLDEN,N.E.

    1999-09-10

    RADIATION DOSIMETRY MEASUREMENTS HAVE BEEN PERFORMED OVER A PERIOD OF MANY YEARS AT THE HIGH FLUX BEAM REACTOR (HFBR) AND THE MEDICAL RESEARCH REACTOR (BMRR) AT BROOKHAVEN NATIONAL LABORATORY TO PROVIDE INFORMATION ON THE ENERGY DISTRIBUTION OF THE NEUTRON FLUX, NEUTRON DOSE RATES, GAMMA-RAY FLUXES AND GAMMA-RAY DOSE RATES. THE MCNP PARTICLE TRANSPORT CODE PROVIDED MONTE CARLO RESULTS TO COMPARE WITH VARIOUS DOSIMETRY MEASUREMENTS PERFORMED AT THE EXPERIMENTAL PORTS, AT THE TREATMENT ROOMS AND IN THE THIMBLES AT BOTH HFBR AND BMRR.

  3. High Flux Metallic Membranes for Hydrogen Recovery and Membrane Reactors

    SciTech Connect (OSTI)

    Buxbaum, Robert

    2010-06-30

    We made and tested over 250 new alloys for use as lower cost, higher flux hydrogen extraction membrane materials. Most of these were intermetallic, or contained significant intermetallic content, particularly based on B2 alloy compositions with at least one refractory component; B2 intermetallics resemble BCC alloys, in structure, but the atoms have relatively fixed positions, with one atom at the corners of the cube, the other at the centers. The target materals we were looking for would contain little or no expensive elements, no strongly toxic or radioactive elements, would have high flux to hydrogen, while being fabricable, brazable, and relatively immune to hydrogen embrittlement and corrosion in operation. The best combination of properties of the membrane materials we developed was, in my opinion, a Pd-coated membrane consisting of V -9 atomic % Pd. This material was relatively cheap, had 5 times the flux of Pd under the same pressure differential, was reasonably easy to fabricate and braze, and not bad in terms of embrittlement. Based on all these factors we project, about 1/3 the cost of Pd, on an area basis for a membrane designed to last 20 years, or 1/15 the cost on a flux basis. Alternatives to this membrane replaced significant fractions of the Pd with Ni and or Co. The cost for these membranes was lower, but so was the flux. We produced successful brazed products from the membrane materials, and made them into flat sheets. We tested, unsuccessfully, several means of fabricating thematerials into tubes, and eventually built a membrane reactor using a new, flat-plate design: a disc and doughnut arrangement, a design that seems well- suited to clean hydrogen production from coal. The membranes and reactor were tested successfully at Western Research. A larger equipment company (Chart Industries) produced similar results using a different flat-plate reactor design. Cost projections of the membrane are shown to be attractive.

  4. Radiation dosimetry at the BNL High Flux Beam Reactor

    SciTech Connect (OSTI)

    Holden, N.E.; Hu, J.P.; Reciniello, R.N.

    1998-02-01

    The HFBR is a heavy water, D{sub 2}O, cooled and moderated reactor with twenty-eight fuel elements containing a maximum of 9.8 kilograms of {sup 235}U. The core is 53 cm high and 48 cm in diameter and has an active volume of 97 liters. The HFBR, which was designed to operate at forty mega-watts, 40 NW, was upgraded to operate at 60 NW. Since 1991, it has operated at 30 MW. In a normal 30 MW operating cycle the HFBR operates 24 hours a day for thirty days, with a six to fourteen day shutdown period for refueling and maintenance work. While most reactors attempts to minimize the escape of neutrons from the core, the HFBR`s D{sub 2}O design allows the thermal neutron flux to peak in the reflector region and maximizes the number of thermal neutrons available to nine horizontal external beams, H-1 to H-9. The HFBR neutron dosimetry effort described here compares measured and calculated energy dependent neutron and gamma ray flux densities and/or dose rates at horizontal beam lines and vertical irradiation thimbles.

  5. Evaluation of selected ex-reactor accidents related to the tritium and medical isotope production mission at the FFTF

    SciTech Connect (OSTI)

    Himes, D.A.

    1997-11-17

    The Fast Flux Test Facility (FFTF) has been proposed as a production facility for tritium and medical isotopes. A range of postulated accidents related to ex-reactor irradiated fuel and target handling were identified and evaluated using new source terms for the higher fuel enrichment and for the tritium and medical isotope targets. In addition, two in-containment sodium spill accidents were re-evaluated to estimate effects of increased fuel enrichment and the presence of the Rapid Retrieval System. Radiological and toxicological consequences of the analyzed accidents were found to be well within applicable risk guidelines.

  6. FFTF (Fast Flux Test Facility) reactor shutdown system reliability reevaluation

    SciTech Connect (OSTI)

    Pierce, B.F.

    1986-07-01

    The reliability analysis of the Fast Flux Test Facility reactor shutdown system was reevaluated. Failure information based on five years of plant operating experience was used to verify original reliability numbers or to establish new ones. Also, system modifications made subsequent to performance of the original analysis were incorporated into the reevaluation. Reliability calculations and sensitivity analyses were performed using a commercially available spreadsheet on a personal computer. The spreadsheet was configured so that future failures could be tracked and compared with expected failures. A number of recommendations resulted from the reevaluation including both increased and decreased surveillance intervals. All recommendations were based on meeting or exceeding existing reliability goals. Considerable cost savings will be incurred upon implementation of the recommendations.

  7. Computer analyses for the design, operation and safety of new isotope production reactors: A technology status review

    SciTech Connect (OSTI)

    Wulff, W.

    1990-01-01

    A review is presented on the currently available technologies for nuclear reactor analyses by computer. The important distinction is made between traditional computer calculation and advanced computer simulation. Simulation needs are defined to support the design, operation, maintenance and safety of isotope production reactors. Existing methods of computer analyses are categorized in accordance with the type of computer involved in their execution: micro, mini, mainframe and supercomputers. Both general and special-purpose computers are discussed. Major computer codes are described, with regard for their use in analyzing isotope production reactors. It has been determined in this review that conventional systems codes (TRAC, RELAP5, RETRAN, etc.) cannot meet four essential conditions for viable reactor simulation: simulation fidelity, on-line interactive operation with convenient graphics, high simulation speed, and at low cost. These conditions can be met by special-purpose computers (such as the AD100 of ADI), which are specifically designed for high-speed simulation of complex systems. The greatest shortcoming of existing systems codes (TRAC, RELAP5) is their mismatch between very high computational efforts and low simulation fidelity. The drift flux formulation (HIPA) is the viable alternative to the complicated two-fluid model. No existing computer code has the capability of accommodating all important processes in the core geometry of isotope production reactors. Experiments are needed (heat transfer measurements) to provide necessary correlations. It is important for the nuclear community, both in government, industry and universities, to begin to take advantage of modern simulation technologies and equipment. 41 refs.

  8. 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.

    2010-12-01

    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

  9. Updated flux information for neutron scattering and irradiation facilities at the BNL High Flux Beam Reactor

    SciTech Connect (OSTI)

    Holden, N.E.; Hu, J.P.; Reciniello, R.N.; Sengupta, S.; Greenwood, L.R.; Farrell, K.

    1997-08-01

    The HFBR is a heavy water, D{sub 2}O, cooled and moderated reactor with twenty-eight fuel elements containing a maximum of 9.8 kilograms of {sup 235}U. While most reactors attempt to minimize the escape of neutrons from the core, the HFBR`s D{sub 2}O design allows the thermal neutron flux to peak in the reflector region and maximizes the number of thermal neutrons available to nine horizontal external beams, H-1 to H-9, used for neutron scattering and capture reactions, supporting physics, chemistry and biology experiments. All horizontal beam tubes were built tangential to the direction of the emerging neutrons, except for the H-2 beam tube, which looks directly at the core and has been used for neutron cross section measurements utilizing fast neutrons and for the TRISTAN fission product studies. In recent years, there have been some beam modifications and new instrumentation introduced at the HFBR. A high resolution neutron powder diffractometer instrument is now operating with a resolution of 5 {times} 10{sup {minus}4} at horizontal beam line H-1. To study scattering from liquid surfaces, a neutron reflection spectrometer was introduced on the CNF beam line at H-9. In the past year, a fourth beam line has been added to the CNF line at H-9. The existing beam plug at the H-6 beam line has recently been removed and a new plug, which will feature super mirrored surfaces, is now being installed. Last year, the vertical beam thimble, V-13, a fixed port filled with thirty year old samples used for HFBR material surveillance studies was replaced by a new thimble and charging station at the core edge creating an irradiation facility to substitute for the original V-13. A neutron dosimetry program has begun to measure and calculate the energy dependent neutron and gamma ray flux densities and/or dose rates at horizontal beam lines and vertical irradiation thimbles.

  10. Modeling and Simulations for the High Flux Isotope Reactor Cycle...

    Office of Scientific and Technical Information (OSTI)

    serve as reference for the design of an LEU fuel for HFIR. ... critical experiments at HFIR, data included in the current HFIR safety analysis report, andor data from ...

  11. Fast Flux Test Facility Reactor Vessel Removal Study

    SciTech Connect (OSTI)

    BOWMAN, B.R.

    2002-10-23

    This study assesses the feasibility of removing the FFTF reactor vessel from its current location in the reactor cavity inside the Containment vessel to a transporter for relocation to a burial pit in the 200 Area.

  12. Level 1 Tornado PRA for the High Flux Beam Reactor

    SciTech Connect (OSTI)

    Bozoki, G.E.; Conrad, C.S.

    1994-05-01

    This report describes a risk analysis primarily directed at providing an estimate for the frequency of tornado induced damage to the core of the High Flux Beam Reactor (HFBR), and thus it constitutes a Level 1 Probabilistic Risk Assessment (PRA) covering tornado induced accident sequences. The basic methodology of the risk analysis was to develop a ``tornado specific`` plant logic model that integrates the internal random hardware failures with failures caused externally by the tornado strike and includes operator errors worsened by the tornado modified environment. The tornado hazard frequency, as well as earlier prepared structural and equipment fragility data, were used as input data to the model. To keep modeling/calculational complexity as simple as reasonable a ``bounding`` type, slightly conservative, approach was applied. By a thorough screening process a single dominant initiating event was selected as a representative initiator, defined as: ``Tornado Induced Loss of Offsite Power.`` The frequency of this initiator was determined to be 6.37E-5/year. The safety response of the HFBR facility resulted in a total Conditional Core Damage Probability of .621. Thus, the point estimate of the HFBR`s Tornado Induced Core Damage Frequency (CDF) was found to be: (CDF){sub Tornado} = 3.96E-5/year. This value represents only 7.8% of the internal CDF and thus is considered to be a small contribution to the overall facility risk expressed in terms of total Core Damage Frequency. In addition to providing the estimate of (CDF){sub Tornado}, the report documents, the relative importance of various tornado induced system, component, and operator failures that contribute most to (CDF){sub Tornado}.

  13. Fast flux test reactor fuel canister. (Journal Article) | SciTech...

    Office of Scientific and Technical Information (OSTI)

    Fast flux test reactor fuel canister. Citation Details ... 952779 Report Number(s): SAND2004-2604J TRN: US0902577 DOE Contract Number: AC04-94AL85000 Resource Type: Journal ...

  14. Application of the Isotope Ratio Method to a Boiling Water Reactor

    SciTech Connect (OSTI)

    Frank, Douglas P.; Gerlach, David C.; Gesh, Christopher J.; Hurley, David E.; Meriwether, George H.; Mitchell, Mark R.; Reid, Bruce D.

    2010-08-11

    The isotope ratio method is a technique for estimating the energy or plutonium production in a fission reactor by measuring isotope ratios in non-fuel reactor components. The isotope ratios in these components can then be directly related to the cumulative energy production with standard reactor modeling methods. All reactor materials contain trace elemental impurities at parts per million levels, and the isotopes of these elements are transmuted by neutron irradiation in a predictable manner. While measuring the change in a particular isotopes concentration is possible, it is difficult to correlate to energy production because the initial concentration of that element may not be accurately known. However, if the ratio of two isotopes of the same element can be measured, the energy production can then be determined without knowing the absolute concentration of that impurity since the initial natural ratio is known. This is the fundamental principle underlying the isotope ratio method. Extremely sensitive mass-spectrometric methods are currently available that allow accurate measurements of the impurity isotope ratios in samples. Additionally, indicator elements with stable activation products have been identified so that their post-irradiation isotope ratios remain constant. This method has been successfully demonstrated on graphite-moderated reactors. Graphite reactors are particularly well-suited to such analyses since the graphite moderator is resident in the fueled region of the core for the entire period of operation. Applying this method to other reactor types is more difficult since the resident portions of the reactor available for sampling are either outside the fueled region of the core or structural components of individual fuel assemblies. The goal of this research is to show that the isotope ratio method can produce meaningful results for light water-moderated power reactors. In this work, we use the isotope ratio method to estimate the energy

  15. Nested reactor chamber and operation for Hg-196 isotope separation process

    DOE Patents [OSTI]

    Grossman, Mark W.

    1991-01-01

    The present invention is directed to an apparatus for use in .sup.196 Hg separation and its method of operation. Specifically, the present invention is directed to a nested reactor chamber useful for .sup.196 Hg isotope separation reactions avoiding the photon starved condition commonly encountered in coaxial reactor systems.

  16. Nested reactor chamber and operation for Hg-196 isotope separation process

    DOE Patents [OSTI]

    Grossman, M.W.

    1991-10-08

    The present invention is directed to an apparatus for use in [sup 196]Hg separation and its method of operation. Specifically, the present invention is directed to a nested reactor chamber useful for [sup 196]Hg isotope separation reactions avoiding the photon starved condition commonly encountered in coaxial reactor systems. 6 figures.

  17. EIS-0291: High Flux Beam Reactor (HFBR) Transition Project at the Brookhaven National Laboratory, Upton, New York

    Broader source: Energy.gov [DOE]

    The EIS evaluates the range of reasonable alternatives and their impacts regarding the future management of the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL).

  18. LETTER REPORT INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT FAN HOUSE, BUILDING 704 BNL

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-10-22

    5098-LR-01-0 -LETTER REPORT INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT FAN HOUSE, BUILDING 704 BROOKHAVEN NATIONAL LABORATORY

  19. INDEPENDENT VERIFICATION SURVEY OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT OUTSIDE AREAS BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-12-15

    5098-SR-03-0 FINAL REPORT- INDEPENDENT VERIFICATION SURVEY OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT OUTSIDE AREAS, BROOKHAVEN NATIONAL LABORATORY

  20. Long-Term Assessment of Isotopic Exchange of Carbon Dioxide in a Subalpine Forest (Niwot Ridge AmeriFlux Site)

    SciTech Connect (OSTI)

    Bowling, David

    2014-12-31

    In 2005 we began a long-term measurement program of CO{sub 2} and its stable isotopes at the Niwot Ridge AmeriFlux site. Measurements are ongoing.

  1. Neutron-flux profile monitor for use in a fission reactor

    DOE Patents [OSTI]

    Kopp, M.K.; Valentine, K.H.

    1981-09-15

    A neutron flux monitor is provided which consists of a plurality of fission counters arranged as spaced-apart point detectors along a delay line. As a fission event occurs in any one of the counters, two delayed current pulses are generated at the output of the delay line. The time separation of the pulses identifies the counter in which the particular fission event occurred. Neutron flux profiles of reactor cores can be more accurately measured as a result.

  2. Apparatus for high flux photocatalytic pollution control using a rotating fluidized bed reactor

    DOE Patents [OSTI]

    Tabatabaie-Raissi, Ali; Muradov, Nazim Z.; Martin, Eric

    2003-06-24

    An apparatus based on optimizing photoprocess energetics by decoupling of the process energy efficiency from the DRE for target contaminants. The technique is applicable to both low- and high-flux photoreactor design and scale-up. An apparatus for high-flux photocatalytic pollution control is based on the implementation of multifunctional metal oxide aerogels and other media in conjunction with a novel rotating fluidized particle bed reactor.

  3. REACTOR

    DOE Patents [OSTI]

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

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

    DOE Patents [OSTI]

    Grossman, Mark W.; Speer, Richard

    1992-01-01

    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.

  5. OECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic Compositions for World Reactor Designs

    SciTech Connect (OSTI)

    Gauld, Ian C; Sly, Nicholas C; Michel-Sendis, Franco

    2014-01-01

    Experimental data on the isotopic concentrations in irradiated nuclear fuel represent one of the primary methods for validating computational methods and nuclear data used for reactor and spent fuel depletion simulations that support nuclear fuel cycle safety and safeguards programs. Measurement data have previously not been available to users in a centralized or searchable format, and the majority of accessible information has been, for the most part, limited to light-water-reactor designs. This paper describes a recent initiative to compile spent fuel benchmark data for additional reactor designs used throughout the world that can be used to validate computer model simulations that support nuclear energy and nuclear safeguards missions. Experimental benchmark data have been expanded to include VVER-440, VVER-1000, RBMK, graphite moderated MAGNOX, gas cooled AGR, and several heavy-water moderated CANDU reactor designs. Additional experimental data for pressurized light water and boiling water reactor fuels has also been compiled for modern assembly designs and more extensive isotopic measurements. These data are being compiled and uploaded to a recently revised structured and searchable database, SFCOMPO, to provide the nuclear analysis community with a centrally-accessible resource of spent fuel compositions that can be used to benchmark computer codes, models, and nuclear data. The current version of SFCOMPO contains data for eight reactor designs, 20 fuel assembly designs, more than 550 spent fuel samples, and measured isotopic data for about 80 nuclides.

  6. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    SciTech Connect (OSTI)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  7. 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. Sulaberidze, G. A.; Alekseev, P. N.; Dudnikov, A. A.; Nevinitsa, V. A. Proselkov, V. N.; Chibinyaev, A. V.

    2012-12-15

    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.

  8. Measurement of the reactor antineutrino flux and spectrum at Daya Bay

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

    D. E. Jaffe; Bishai, M; Diwan, M.; Gill, R.; Hackenburg, R. W.; Hans, S.; Hu, L. M.; Jaffe, D. E.; Kettell, S. H.; Tang, W.; et al

    2016-02-12

    This Letter reports a measurement of the flux and energy spectrum of electron antineutrinos from six 2.9~GWth nuclear reactors with six detectors deployed in two near (effective baselines 512~m and 561~m) and one far (1,579 m) underground experimental halls in the Daya Bay experiment. Using 217 days of data, 296,721 and 41,589 inverse beta decay (IBD) candidates were detected in the near and far halls, respectively. The measured IBD yield is (1.55 ± 0.04) × 10–18 cm2/GW/day or (5.92 ± 0.14) × 10–43 cm2/fission. This flux measurement is consistent with previous short-baseline reactor antineutrino experiments and is 0.946 ± 0.022more » (0.991 ± 0.023) relative to the flux predicted with the Huber+Mueller (ILL+Vogel) fissile antineutrino model. The measured IBD positron energy spectrum deviates from both spectral predictions by more than 2σ over the full energy range with a local significance of up to ~4σ between 4-6 MeV. Furthermore, a reactor antineutrino spectrum of IBD reactions is extracted from the measured positron energy spectrum for model-independent predictions.« less

  9. TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 2 DF WASTE LINE REMOVAL, BNL

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-07-09

    5098-SR-02-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 2 DF WASTE LINE REMOVAL, BROOKHAVEN NATIONAL LABORATORY

  10. Fukushima Daiichi reactor source term attribution using cesium isotope ratios from contaminated environmental samples

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

    Snow, Mathew S.; Snyder, Darin C.; Delmore, James E.

    2016-01-18

    Source term attribution of environmental contamination following the Fukushima Daiichi Nuclear Power Plant (FDNPP) disaster is complicated by a large number of possible similar emission source terms (e.g. FDNPP reactor cores 1–3 and spent fuel ponds 1–4). Cesium isotopic analyses can be utilized to discriminate between environmental contamination from different FDNPP source terms and, if samples are sufficiently temporally resolved, potentially provide insights into the extent of reactor core damage at a given time. Rice, soil, mushroom, and soybean samples taken 100–250 km from the FDNPP site were dissolved using microwave digestion. Radiocesium was extracted and purified using two sequentialmore » ammonium molybdophosphate-polyacrylonitrile columns, following which 135Cs/137Cs isotope ratios were measured using thermal ionization mass spectrometry (TIMS). Results were compared with data reported previously from locations to the northwest of FDNPP and 30 km to the south of FDNPP. 135Cs/137Cs isotope ratios from samples 100–250 km to the southwest of the FDNPP site show a consistent value of 0.376 ± 0.008. 135Cs/137Cs versus 134Cs/137Cs correlation plots suggest that radiocesium to the southwest is derived from a mixture of FDNPP reactor cores 1, 2, and 3. Conclusions from the cesium isotopic data are in agreement with those derived independently based upon the event chronology combined with meteorological conditions at the time of the disaster. In conclusion, cesium isotopic analyses provide a powerful tool for source term discrimination of environmental radiocesium contamination at the FDNPP site. For higher precision source term attribution and forensic determination of the FDNPP core conditions based upon cesium, analyses of a larger number of samples from locations to the north and south of the FDNPP site (particularly time-resolved air filter samples) are needed. Published in 2016. This article is a U.S. Government work and is in the public domain

  11. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOE Patents [OSTI]

    Lasche, G.P.

    1983-09-29

    The invention is a laser or particle-beam-driven fusion reactor system which takes maximum advantage of both the very short pulsed nature of the energy release of inertial confinement fusion (ICF) and the very small volumes within which the thermonuclear burn takes place. The pulsed nature of ICF permits dynamic direct energy conversion schemes such as magnetohydrodynamic (MHD) generation and magnetic flux compression; the small volumes permit very compact blanket geometries. By fully exploiting these characteristics of ICF, it is possible to design a fusion reactor with exceptionally high power density, high net electric efficiency, and low neutron-induced radioactivity. The invention includes a compact blanket design and method and apparatus for obtaining energy utilizing the compact blanket.

  12. The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

    SciTech Connect (OSTI)

    Nevinitsa, V. A. Dudnikov, A. A.; Blandinskiy, V. Yu.; Balanin, A. L.; Alekseev, P. N.; Titarenko, Yu. E.; Batyaev, V. F.; Pavlov, K. V.; Titarenko, A. Yu.

    2015-12-15

    A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing {sup 233}U from {sup 232}Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

  13. Isotopes

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

    Office of Science » Nuclear Physics » Isotopes Isotopes Isotopes produced at Los Alamos National Laboratory are saving lives, advancing cutting-edge research and keeping the U.S. safe. Get Expertise Eva Birnbaum (505) 665-7167 Email Wolfgang Runde (505) 667-3350 Email Isotope Production and Applications isotopes Isotopes produced at IPF are critical for medical diagnosis and disease treatment. These positron emission tomography images were made possible using isotopes produced at LANL.

  14. Solid tags for identifying failed reactor components

    DOE Patents [OSTI]

    Bunch, Wilbur L.; Schenter, Robert E.

    1987-01-01

    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. Use of LEU in the aqueous homogeneous medical isotope production reactor

    SciTech Connect (OSTI)

    Ball, R.M.

    1997-08-01

    The Medical Isotope Production Reactor (MIPR) is an aqueous solution of uranyl nitrate in water, contained in an aluminum cylinder immersed in a large pool of water which can provide both shielding and a medium for heat exchange. The control rods are inserted at the top through re-entrant thimbles. Provision is made to remove radiolytic gases and recombine emitted hydrogen and oxygen. Small quantities of the solution can be continuously extracted and replaced after passing through selective ion exchange columns, which are used to extract the desired products (fission products), e.g. molybdenum-99. This reactor type is known for its large negative temperature coefficient, the small amount of fuel required for criticality, and the ease of control. Calculation using TWODANT show that a 20% U-235 enriched system, water reflected can be critical with 73 liters of solution.

  16. Reactor production of Thoruim-229

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

    Boll, Rose Ann; Murphy, Karen E.; Denton, David L.; Tamara J. Haverlock; Garland, Marc A.; Mirzadeh, Saed; Hogle, Susan; Owens, Allison

    2016-05-03

    Limited availability of 229Th for clinical applications of 213Bi necessitates investigation of alternative production routes. In reactor production, 229Th is produced from neutron transmutation of 226Ra, 228Ra, 227Ac and 228Th. Here, we evaluate irradiations of 226Ra, 228Ra, and 227Ac targets at the ORNL High Flux Isotope Reactor.

  17. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 1, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    SciTech Connect (OSTI)

    E.M. Harpenau

    2010-12-15

    5098-SR-05-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 1 BROOKHAVEN NATIONAL LABORATORY

  18. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 5, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-11-03

    5098-SR-04-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 5, BROOKHAVEN NATIONAL LABORATORY

  19. Isotopes

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

    Eva Birnbaum (505) 665-7167 Email Wolfgang Runde (505) 667-3350 Email Isotope Production and Applications isotopes Isotopes produced at IPF are critical for medical diagnosis and ...

  20. Potential role of the Fast Flux Test Facility and the advanced test reactor in the U.S. tritium production system

    SciTech Connect (OSTI)

    Dautel, W.A.

    1996-10-01

    The Deparunent of Energy is currently engaged in a dual-track strategy to develop an accelerator and a conunercial light water reactor (CLWR) as potential sources of tritium supply. New analysis of the production capabilities of the Fast Flux Test Facility (FFTF) at the Hanford Site argues for considering its inclusion in the tritium supply,system. The use of the FFTF (alone or together with the Advanced Test Reactor [ATR] at the Idaho National Engineering Laboratory) as an integral part of,a tritium production system would help (1) ensure supply by 2005, (2) provide additional time to resolve institutional and technical issues associated with the- dual-track strategy, and (3) reduce discounted total life-cycle`costs and near-tenn annual expenditures for accelerator-based systems. The FFRF would also provide a way to get an early start.on dispositioning surplus weapons-usable plutonium as well as provide a source of medical isotopes. Challenges Associated With the Dual-Track Strategy The Departinent`s purchase of either a commercial reactor or reactor irradiation services faces challenging institutional issues associated with converting civilian reactors to defense uses. In addition, while the technical capabilities of the individual components of the accelerator have been proven, the entire system needs to be demonstrated and scaled upward to ensure that the components work toge ther 1548 as a complete production system. These challenges create uncertainty over the ability of the du2a-track strategy to provide an assured tritium supply source by 2005. Because the earliest the accelerator could come on line is 2007, it would have to operate at maximum capacity for the first few years to regenerate the reserves lost through radioactive decay aftei 2005.

  1. CO2 and CH4 Surface Flux, Soil Profile Concentrations, and Stable Isotope Composition, Barrow, Alaska, 2012-2013

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

    Curtis, J.B.; Vaughn, L.S.; Torn, M.S.; Conrad, M.S.; Chafe, O.; Bill, M.

    2015-12-31

    In August-October 2012 and June-October 2013, co-located measurements were made of surface CH4 and CO2 flux, soil pore space concentrations and stable isotope compositions of CH4 and CO2, and subsurface temperature and soil moisture. Measurements were made in intensive study site 1 areas A, B, and C, and from the site 0 and AB transects, from high-centered, flat-centered, and low-centered polygons, from the center, edge, and trough of each polygon.

  2. A brief History of Neutron Scattering at the Oak Ridge High Flux...

    Office of Scientific and Technical Information (OSTI)

    A brief History of Neutron Scattering at the Oak Ridge High Flux Isotope Reactor Citation Details In-Document Search Title: A brief History of Neutron Scattering at the Oak Ridge ...

  3. Total absorption spectroscopy study of ?Rb decay: A major contributor to reactor antineutrino spectrum shape [Total absorption spectroscopy study of ?Rb: A major contributor to reactor antineutrino flux

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

    Sonzogni, A.; Zakari-Issoufou, A. -A.; Fallot, M.; Porta, A.; Algora, A.; Tain, J. L.; Valencia, E.; Rice, S.; Bui, V. M.; Cormon, S.; et al

    2015-03-09

    The accurate determination of the emitted reactor antineutrino flux is still a major challenge for actual and future neutrino experiments at reactors, especially after the evidence of a disagreement between the measured antineutrino energy spectrum by Double Chooz, Daya Bay, and Reno and calculated antineutrino spectra obtained from the conversion of the unique integral beta spectra measured at the ILL reactor. Using nuclear data to compute reactor antineutrino spectra may help understanding this bias, with the study of the underlying nuclear physics. Summation calculations allow identifying a list of nuclei that contribute importantly to the antineutrino energy spectra emitted aftermorethe fission of ?,?Pu and ?,?U, and whose beta decay properties might deserve new measurements. Among these nuclei, ?Rb exhausts by itself about 16% of of the antineutrino energy spectrum emitted by Pressurized Water Reactors in the 5 to 8 MeV range. In this Letter, we report new Total Absorption Spectroscopy (TAS) results for this important contributor. The obtained beta feeding from ?Rb shows beta intensity unobserved before in the 4.5 to 5.5 MeV energy region and gives a ground state to ground state branch of 87.5 % 3%. These new data induce a dramatic change in recent summation calculations where a 51% GS to GS branch was considered for ?Rb, increasing the summation antineutrino spectrum in the region nearby the observed bias.The new data still have an important impact on other summation calculations in which more recent data were consideredless

  4. CRAD, Fire 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 2006 Commencement of Operations assessment of the Fire Protection program at the Idaho Accelerated Retrieval Project Phase II.

  5. CRAD, DOE Oversight- 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 DOE independent oversight assessment of the Oak Ridge National Laboratory programs for oversight of its contractors.

  6. Scientific Upgrades at the High Flux Isotope Reactor at Oak Ridge...

    Office of Scientific and Technical Information (OSTI)

    This paper provides a short summary of these projectsmore including their present status and schedule. less Authors: Selby, Douglas L 1 ; Smith, Gregory Scott 1 + Show ...

  7. HFIR (High Flux Isotope Reactor) pressure vessel and structural components materials surveillance program: Supplement 1

    SciTech Connect (OSTI)

    Cheverton, R.D.; McGinty, D.M.; McWherter, J.R.; Nanstad, R.K.

    1987-10-01

    Extending the life of the HFIR vessel by the proposed 10 effective full-power years is contingent upon a continuation of the materials surveillance program and the application of hydrostatic proof testing. As a part of the surveillance program, Charpy V-notch (CVN) specimens of shell, weld and nozzle materials are installed adjacent to the inner surface of the vessel and are removed periodically for testing to determine the radiation-induced increase in the nil-ductility transition temperature. Hydro testing is conducted to prove that a critical combination of flaw size, stress and fracture toughness does not exist. Information from the materials surveillance program is used in a fracture mechanics analysis to confirm that the hydro-test pressure being applied is appropriate for the desired life extension of the vessel. This report specifies (1) the number, type, location and schedule for removal-testing of the CVN specimens for the continuing materials surveillance program, and (2) the procedures and test conditions for the hydro test.

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

    SciTech Connect (OSTI)

    Houser, M.M.

    1995-08-01

    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.

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

    SciTech Connect (OSTI)

    Ilas, Germina; Gauld, Ian C

    2011-01-01

    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.

  10. Total absorption spectroscopy study of ?Rb decay: A major contributor to reactor antineutrino spectrum shape [Total absorption spectroscopy study of ?Rb: A major contributor to reactor antineutrino flux

    SciTech Connect (OSTI)

    Sonzogni, A.; Zakari-Issoufou, A. -A.; Fallot, M.; Porta, A.; Algora, A.; Tain, J. L.; Valencia, E.; Rice, S.; Bui, V. M.; Cormon, S.; Estienne, M.; Agramunt, J.; Aysto, J.; Bowry, M.; Briz Monago, J. A.; Caballero-Folch, R.; Cano-Ott, D.; Cucoanes, A.; Eloma, V.; Estvez, E.; Farrelly, G. F.; Garcia, A.; Gelletly, W.; Gomez-Hornillos, M. B.; Gorlychev, V.; Hakala, J.; Jokinen, A.; Jordan, M. D.; Kankainen, A.; Kondev, F. G.; Martinez, T.; Mendoza, E.; Molina, F.; Moore, I.; Perez, A.; Podolyak, Zs.; Penttil, H.; Regan, P. H.; Shiba, T.; Rissanen, J.; Rubio, B.; Weber, C.

    2015-03-09

    The accurate determination of the emitted reactor antineutrino flux is still a major challenge for actual and future neutrino experiments at reactors, especially after the evidence of a disagreement between the measured antineutrino energy spectrum by Double Chooz, Daya Bay, and Reno and calculated antineutrino spectra obtained from the conversion of the unique integral beta spectra measured at the ILL reactor. Using nuclear data to compute reactor antineutrino spectra may help understanding this bias, with the study of the underlying nuclear physics. Summation calculations allow identifying a list of nuclei that contribute importantly to the antineutrino energy spectra emitted after the fission of ?,?Pu and ?,?U, and whose beta decay properties might deserve new measurements. Among these nuclei, ?Rb exhausts by itself about 16% of of the antineutrino energy spectrum emitted by Pressurized Water Reactors in the 5 to 8 MeV range. In this Letter, we report new Total Absorption Spectroscopy (TAS) results for this important contributor. The obtained beta feeding from ?Rb shows beta intensity unobserved before in the 4.5 to 5.5 MeV energy region and gives a ground state to ground state branch of 87.5 % 3%. These new data induce a dramatic change in recent summation calculations where a 51% GS to GS branch was considered for ?Rb, increasing the summation antineutrino spectrum in the region nearby the observed bias.The new data still have an important impact on other summation calculations in which more recent data were considered

  11. High heat flux testing of HIP bonded DS-Cu/316SS first wall panel for fusion experimental reactors

    SciTech Connect (OSTI)

    Hatano, Toshihisa; Sato, Kazuyoshi; Dairaku, Masayuki

    1996-12-31

    A shielding blanket design in a fusion reactor such as ITER has been proposed to be a modulator structure integrated with the first wall. In terms of the fabrication, HIP (Hot Isostatic Pressing) method has been proposed for the joining of dispersion strengthened copper (DS-Cu) and type 316L stainless steel (SS316L) at FW. High heat flux tests of HIP bonded DS-Cu/SS316L first wall panel were performed at particle Beam Engineering Facility in JAERI to investigate its thermo-mechanical performance. After four campaigns of high heat flux testing, the FW panel was cut to observe the HIP bonded interface and heated surface of DS-Cu. Though melting of DS-Cu surface was observed, there were no cracks at the HIP bonded interface. 2 refs., 11 figs., 1 tab.

  12. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    SciTech Connect (OSTI)

    Lasche, G.P.

    1988-04-05

    A method for recovering energy in an inertial confinement fusion reactor having a reactor chamber and a sphere forming means positioned above an opening in the reactor chamber is described, comprising: embedding a fusion target fuel capsule having a predetermined yield in the center of a hollow solid lithium tube and subsequently embedding the hollow solid lithium tube in a liquid lithium medium; using the sphere forming means for forming the liquid lithium into a spherical shaped liquid lithium mass having a diameter smaller than the length of the hollow solid lithium tube with the hollow solid lithium tube being positioned along a diameter of the spherical shaped mass, providing the spherical shaped liquid lithium mass with the fusion fuel target capsule and hollow solid lithium tube therein as a freestanding liquid lithium shaped spherical shaped mass without any external means for maintaining the spherical shape by dropping the liquid lithium spherical shaped mass from the sphere forming means into the reactor chamber; producing a magnetic field in the reactor chamber; imploding the target capsule in the reactor chamber to produce fusion energy; absorbing fusion energy in the liquid lithium spherical shaped mass to convert substantially all the fusion energy to shock induced kinetic energy of the liquid lithium spherical shaped mass which expands the liquid lithium spherical shaped mass; and compressing the magnetic field by expansion of the liquid lithium spherical shaped mass and recovering useful energy.

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

    SciTech Connect (OSTI)

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

    2010-08-19

    The High Flux Reactor (RHF) of the Laue Langevin Institute (ILL) based in Grenoble, France is a research reactor designed primarily for neutron beam experiments for fundamental science. It delivers one of the most intense neutron fluxes worldwide, with an unperturbed thermal neutron flux of 1.5 x 10{sup 15} n/cm{sup 2}/s in its reflector. The reactor has been conceived to operate at a nuclear power of 57 MW but currently operates at 52 MW. The reactor currently uses a Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most worldwide research and test reactors have already started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on a mixture of uranium and molybdenum (UMo) is expected to allow the conversion of compact high performance reactors like the RHF. This report presents the results of reactor design, performance and steady state safety analyses for conversion of the RHF from the use of HEU fuel to the use of UMo LEU fuel. The objective of this work was to show that is feasible, under a set of manufacturing assumptions, to design a new RHF fuel element that could safely replace the HEU element currently used. The new proposed design has been developed to maximize performance, minimize changes and preserve strong safety margins. Neutronics and thermal-hydraulics models of the RHF have been developed and qualified by benchmark against experiments and/or against other codes and models. The models developed were then used to evaluate the RHF performance if LEU UMo were to replace the current HEU fuel 'meat' without any geometric change to the fuel plates. Results of these direct replacement analyses have shown a significant degradation of the RHF performance, in terms of both neutron flux and cycle length

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

    DOE Patents [OSTI]

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

    1982-08-19

    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.

  15. High Specific Activity Sn-117m by Post Irradiation Isotope Separation

    SciTech Connect (OSTI)

    DAuria, John

    2015-04-16

    ElectroMagnetic Isotope Separation (EMIS) is used in the production of enriched stable isotopes. We demonstrated the feasibility of using EMIS to produce medium Specific Activity 117mSm using high purity 116Sn target material irradiated in a high flux reactor.

  16. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOE Patents [OSTI]

    Lasche, G.P.

    1987-02-20

    A high-power-density-laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems. 25 figs.

  17. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOE Patents [OSTI]

    Lasche, George P.

    1988-01-01

    A high-power-density laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems.

  18. Progress in the Use of Isotopes: The Atomic Triad - Reactors, Radioisotopes and Radiation

    DOE R&D Accomplishments [OSTI]

    Libby, W. F.

    1958-08-04

    Recent years have seen a substantial growth in the use of isotopes in medicine, agriculture, and industry: up to the minute information on the production and use of isotopes in the U.S. is presented. The application of radioisotopes to industrial processes and manufacturing operations has expanded more rapidly than any one except its most ardent advocates expected. New uses and new users are numerous. The adoption by industry of low level counting techniques which make possible the use of carbon-14 and tritium in the control of industrial processes and in certain exploratory and research problems is perhaps most promising of current developments. The latest information on savings to industry will be presented. The medical application of isotopes has continued to develop at a rapid pace. The current trend appears to be in the direction of improvements in technique and the substitution of more effective isotopes for those presently in use. Potential and actual benefits accruing from the use of isotopes in agriculture are reviewed. The various methods of production of radioisotopes are discussed. Not only the present methods but also interesting new possibilities are covered. Although isotopes are but one of the many peaceful uses of the atom, it is the first to pay its way. (auth)

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

    SciTech Connect (OSTI)

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

    2015-12-15

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

  20. An experimental study of external reactor vessel cooling strategy on the critical heat flux using the graphene oxide nano-fluid

    SciTech Connect (OSTI)

    Park, S. D.; Lee, S. W.; Kang, S.; Kim, S. M.; Seo, H.; Bang, I. C.

    2012-07-01

    External reactor vessel cooling (ERVC) for in-vessel retention (IVR) of corium as a key severe accident management strategy can be achieved by flooding the reactor cavity during a severe accident. In this accident mitigation strategy, the decay heat removal capability depends on whether the imposed heat flux exceeds critical heat flux (CHF). To provide sufficient cooling for high-power reactors such as APR1400, there have been some R and D efforts to use the reactor vessel with micro-porous coating and nano-fluids boiling-induced coating. The dispersion stability of graphene-oxide nano-fluid in the chemical conditions of flooding water that includes boric acid, lithium hydroxide (LiOH) and tri-sodium phosphate (TSP) was checked in terms of surface charge or zeta potential before the CHF experiments. Results showed that graphene-oxide nano-fluids were very stable under ERVC environment. The critical heat flux (CHF) on the reactor vessel external wall was measured using the small scale two-dimensional slide test section. The radius of the curvature is 0.1 m. The dimension of each part in the facility simulated the APR-1400. The heater was designed to produce the different heat flux. The magnitude of heat flux follows the one of the APR-1400 when the severe accident occurred. All tests were conducted under inlet subcooling 10 K. Graphene-oxide nano-fluids (concentration: 10 -4 V%) enhanced CHF limits up to about 20% at mass flux 50 kg/m{sup 2}s and 100 kg/m{sup 2}s in comparison with the results of the distilled water at same test condition. (authors)

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

    DOE Patents [OSTI]

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

    1982-08-19

    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.

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

    DOE Patents [OSTI]

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

    1982-08-19

    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.

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

    DOE Patents [OSTI]

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

    1982-08-19

    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.

  4. Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

    SciTech Connect (OSTI)

    Not Available

    1992-07-01

    Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

  5. Neutronics Conversion Analyses of the Laue-Langevin Institute (ILL) High Flux Reactor (RHF)

    SciTech Connect (OSTI)

    Bergeron, A.; Dionne, B.; Calzavara, Y.

    2014-09-30

    The following report describes the neutronics results obtained with the MCNP model of the RHF U7Mo LEU reference design that has been established in 2010 during the feasibility analysis. This work constitutes a complete and detailed neutronics analysis of that LEU design using models that have been significantly improved since 2010 and the release of the feasibility report. When possible, the credibility of the neutronics model is tested by comparing the HEU model results with experimental data or other codes calculations results. The results obtained with the LEU model are systematically compared to the HEU model. The changes applied to the neutronics model lead to better comparisons with experimental data or improved the calculation efficiency but do not challenge the conclusion of the feasibility analysis. If the U7Mo fuel is commercially available, not cost prohibitive, a back-end solution is established and if it is possible to manufacture the proposed element, neutronics analyses show that the performance of the reactor would not be challenged by the conversion to LEU fuel.

  6. Multi-scale Control and Enhancement of Reactor Boiling Heat Flux by Reagents and Nanoparticles

    SciTech Connect (OSTI)

    Manglik, R M; Athavale, A; Kalaikadal, D S; Deodhar, A; Verma, U

    2011-09-02

    The phenomenological characterization of the use of non-invasive and passive techniques to enhance the boiling heat transfer in water has been carried out in this extended study. It provides fundamental enhanced heat transfer data for nucleate boiling and discusses the associated physics with the aim of addressing future and next-generation reactor thermal-hydraulic management. It essentially addresses the hypothesis that in phase-change processes during boiling, the primary mechanisms can be related to the liquid-vapor interfacial tension and surface wetting at the solidliquid interface. These interfacial characteristics can be significantly altered and decoupled by introducing small quantities of additives in water, such as surface-active polymers, surfactants, and nanoparticles. The changes are fundamentally caused at a molecular-scale by the relative bulk molecular dynamics and adsorption-desorption of the additive at the liquid-vapor interface, and its physisorption and electrokinetics at the liquid-solid interface. At the micro-scale, the transient transport mechanisms at the solid-liquid-vapor interface during nucleation and bubblegrowth can be attributed to thin-film spreading, surface-micro-cavity activation, and micro-layer evaporation. Furthermore at the macro-scale, the heat transport is in turn governed by the bubble growth and distribution, macro-layer heat transfer, bubble dynamics (bubble coalescence, collapse, break-up, and translation), and liquid rheology. Some of these behaviors and processes are measured and characterized in this study, the outcomes of which advance the concomitant fundamental physics, as well as provide insights for developing control strategies for the molecular-scale manipulation of interfacial tension and surface wetting in boiling by means of polymeric reagents, surfactants, and other soluble surface-active additives.

  7. Analysis of the Reactor Position Independent Monitor (PIM) Diagnostic

    SciTech Connect (OSTI)

    Hayes-Sterbenz, Anna Catherine

    2014-07-17

    In this note I analyze the physics determining the proposed reactor position independent monitor (PIM), which is the ratio (240Pu/239Pu)1/3 × (135Cs/137Cs)1/2. The PIM ratios in any reactor fuel is shown to increase monotonically with the time over which the fuel is irradiated. This is because the Cs ratio determines the neutron flux, while the Pu isotopic ratio is determined by the flux times the irradiation time. If the irradiation time for all fuel rods across the reactor is fixed, the PIM ratio is approximately constant in all rods. However, no information can be extracted from the PIM ratio on Pu isotopics unless both the flux (or Cs ratio) and the irradiation time (from, say, Ru isotopics) are known separately, i.e., the PIM ratio is not a fundamental parameter of any reactor. Thus, unless the PIM ratio has been measured for the specific fuel under interrogation, no information can be deduced from measurements or reactor simulations of PIM ratios in different fuel from the same reactor. However, if a PIM measurement has been in one spent fuel rod from a given reactor, all other rods that are known to have been in the reactor for the same irradiation period can be assumed to have approximately the same PIM ratio.

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

    SciTech Connect (OSTI)

    Rosenthal, Murray Wilford

    2009-08-01

    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.

  9. Final Report Independent Verification Survey of the High Flux Beam Reactor, Building 802 Fan House Brookhaven National Laboratory Upton, New York

    SciTech Connect (OSTI)

    Harpeneau, Evan M.

    2011-06-24

    On May 9, 2011, ORISE conducted verification survey activities including scans, sampling, and the collection of smears of the remaining soils and off-gas pipe associated with the 802 Fan House within the HFBR (High Flux Beam Reactor) Complex at BNL. ORISE is of the opinion, based on independent scan and sample results obtained during verification activities at the HFBR 802 Fan House, that the FSS (final status survey) unit meets the applicable site cleanup objectives established for as left radiological conditions.

  10. REACTOR

    DOE Patents [OSTI]

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  11. Five years operating experience at the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Baumhardt, R. J.; Bechtold, R. A.

    1987-04-01

    The Fast Flux Test Facility (FFTF) is a 400 Mw(t), loop-type, sodium-cooled, fast neutron reactor. It is operated by the Westinghouse Hanford Company for the United States Department of Energy at Richland, Washington. The FFTF is a multipurpose test reactor used to irradiate fuels and materials for programs such as Liquid Metal Reactor (LMR) research, fusion research, space power systems, isotope production and international research. FFTF is also used for testing concepts to be used in Advanced Reactors which will be designed to maximize passive safety features and not require complex shutdown systems to assure safe shutdown and heat removal. The FFTF also provides experience in the operation and maintenance of a reactor having prototypic components and systems typical of large LMR (LMFBR) power plants. The 5 year operational performance of the FFTF reactor is discussed in this report. 6 refs., 10 figs., 2 tabs.

  12. REACTOR

    DOE Patents [OSTI]

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  13. REACTORS

    DOE Patents [OSTI]

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  14. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    SciTech Connect (OSTI)

    Slater, C.O.

    1990-07-01

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

  15. A preliminary assessment of the effects of heat flux distribution and penetration on the creep rupture of a reactor vessel lower head

    SciTech Connect (OSTI)

    Chu, T.Y.; Bentz, J.; Simpson, R.; Witt, R.

    1997-02-01

    The objective of the Lower Head Failure (LHF) Experiment Program is to experimentally investigate and characterize the failure of the reactor vessel lower head due to thermal and pressure loads under severe accident conditions. The experiment is performed using 1/5-scale models of a typical PWR pressure vessel. Experiments are performed for various internal pressure and imposed heat flux distributions with and without instrumentation guide tube penetrations. The experimental program is complemented by a modest modeling program based on the application of vessel creep rupture codes developed in the TMI Vessel Investigation Project. The first three experiments under the LHF program investigated the creep rupture of simulated reactor pressure vessels without penetrations. The heat flux distributions for the three experiments are uniform (LHF-1), center-peaked (LHF-2), and side-peaked (LHF-3), respectively. For all the experiments, appreciable vessel deformation was observed to initiate at vessel wall temperatures above 900K and the vessel typically failed at approximately 1000K. The size of failure was always observed to be smaller than the heated region. For experiments with non-uniform heat flux distributions, failure typically occurs in the region of peak temperature. A brief discussion of the effect of penetration is also presented.

  16. Reactor

    DOE Patents [OSTI]

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  17. Type A verification report for the high flux beam reactor stack and grounds, Brookhaven National Laboratory, Upton, New York

    SciTech Connect (OSTI)

    Harpenau, Evan M.

    2012-01-13

    The U.S. Department of Energy (DOE) Order 458.1 requires independent verification (IV) of DOE cleanup projects (DOE 2011). The Oak Ridge Institute for Science and Education (ORISE) has been designated as the responsible organization for IV of the High Flux Beam Reactor (HFBR) Stack and Grounds area at Brookhaven National Laboratory (BNL) in Upton, New York. The IV evaluation may consist of an in-process inspection with document and data reviews (Type A Verification) or a confirmatory survey of the site (Type B Verification). DOE and ORISE determined that a Type A verification of the documents and data for the HFBR Stack and Grounds: Survey Units (SU) 6, 7, and 8 was appropriate based on the initial survey unit classification, the walkover surveys, and the final analytical results provided by the Brookhaven Science Associates (BSA). The HFBR Stack and Grounds surveys began in June 2011 and were completed in September 2011. Survey activities by BSA included gamma walkover scans and sampling of the as-left soils in accordance with the BSA Work Procedure (BNL 2010a). The Field Sampling Plan - Stack and Remaining HFBR Outside Areas (FSP) stated that gamma walk-over surveys would be conducted with a bare sodium iodide (NaI) detector, and a collimated detector would be used to check areas with elevated count rates to locate the source of the high readings (BNL 2010b). BSA used the Mult- Agency Radiation Survey and Site Investigation Manual (MARSSIM) principles for determining the classifications of each survey unit. Therefore, SUs 6 and 7 were identified as Class 1 and SU 8 was deemed Class 2 (BNL 2010b). Gamma walkover surveys of SUs 6, 7, and 8 were completed using a 2?2 NaI detector coupled to a data-logger with a global positioning system (GPS). The 100% scan surveys conducted prior to the final status survey (FSS) sampling identified two general soil areas and two isolated soil locations with elevated radioactivity. The general areas of elevated activity identified

  18. Analysis of palladium coatings to remove hydrogen isotopes from zirconium fuel rods in Canada deuterium uranium-pressurized heavy water reactors; Thermal and neutron diffusion effects

    SciTech Connect (OSTI)

    Stokes, C.L.; Buxbaum, R.E. )

    1992-05-01

    This paper reports that, in pressurized heavy water nuclear reactors of the type standardly used in Canada (Canada deuterium uranium-pressurized heavy water reactors), the zirconium alloy pressure tubes of the core absorb deuterium produced by corrosion reactions. This deuterium weakens the tubes through hydrogen embrittlement. Thin palladium coatings on the outside of the zirconium are analyzed as a method for deuterium removal. This coating is expected to catalyze the reaction D{sub 2} + 1/2O{sub 2} {r reversible} D{sub 2}O when O{sub 2} is added to the annular (insulating) gas in the tubes. Major reductions in the deuterium concentration and, hence, hydrogen embrittlement are predicted. Potential problems such as plating the tube geometry, neutron absorption, catalyst deactivation, radioactive waste production, and oxygen corrosion are shown to be manageable. Also, a simple set of equations are derived to calculate the effect on diffusion caused by neutron interactions. Based on calculations of ordinary and neutron flux induced diffusion, a palladium coating of 1 {times} 10{sup {minus}6} m is recommended. This would cost approximately $60,000 per reactor unit and should more than double reactor lifetime. Similar coatings and similar interdiffusion calculations might have broad applications.

  19. TYPE A VERIFICATION REPORT FOR THE HIGH FLUX BEAM REACTOR STACK AND GROUNDS, BROOKHAVEN NATIONAL LABORATORY, UPTON, NEW YORK DCN 5098-SR-08-0

    SciTech Connect (OSTI)

    Evan Harpenau

    2011-11-30

    The U.S. Department of Energy (DOE) Order 458.1 requires independent verification (IV) of DOE cleanup projects (DOE 2011). The Oak Ridge Institute for Science and Education (ORISE) has been designated as the responsible organization for IV of the High Flux Beam Reactor (HFBR) Stack and Grounds area at Brookhaven National Laboratory (BNL) in Upton, New York. The IV evaluation may consist of an in-process inspection with document and data reviews (Type A Verification) or a confirmatory survey of the site (Type B Verification). DOE and ORISE determined that a Type A verification of the documents and data for the HFBR Stack and Grounds: Survey Units (SU) 6, 7, and 8 was appropriate based on the initial survey unit classification, the walkover surveys, and the final analytical results provided by the Brookhaven Science Associates (BSA).

  20. Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

    SciTech Connect (OSTI)

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; Martinez, Patrick; Keller, Russ; Stanley, Floyd; Spencer, Khalil; Thomas, Mariam; Xu, Ning; Schappert, Michael; Fulwyler, James

    2015-10-26

    This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

  1. Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

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

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; Martinez, Patrick; Keller, Russ; Stanley, Floyd; Spencer, Khalil; Thomas, Mariam; Xu, Ning; Schappert, Michael; et al

    2015-10-26

    This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

  2. Investigation of parameters of interaction of hydrogen isotopes with liquid lithium and lithium capillary-porous system under reactor irradiation

    SciTech Connect (OSTI)

    Tazhibayeva, I. L. Kulsartov, T. V.; Gordienko, Yu. N.; Zaurbekova, Zh. A.; Ponkratov, Yu. V.; Barsukov, N. I.; Tulubayev, Ye. Yu.; Baklanov, V. V.; Gnyrya, V. S.; Kenzhin, Ye. A.

    2015-12-15

    In this study, the effect of reactor irradiation on the processes of interaction of hydrogen with liquid lithium and a lithium capillary-porous system (CPS) is considered. The experiments are carried out by the gas-absorption method with use of a specially designed ampoule device. The results of investigation of the interaction of hydrogen with liquid lithium and a lithium CPS under conditions of reactor irradiation are described; namely, these are the temperature dependences of the rate constant for the interaction of hydrogen with liquid lithium at different reactor powers, the activation energies of the processes, and the pre-exponential factor in the Arrhenius dependence. The effect of increasing absorption of hydrogen by the samples under investigation as a result of the reactor irradiation is fixed. The effect can be explained by increasing mobility of hydrogen in liquid lithium due to hot spots in lithium bulk and the interaction of helium and tritium ions (formed as a result of the nuclear reaction of {sup 6}Li with neutron) with a surface hydride film.

  3. ``Sleeping reactor`` irradiations: Shutdown reactor determination of short-lived activation products

    SciTech Connect (OSTI)

    Jerde, E.A.; Glasgow, D.C.

    1998-09-01

    At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux ({phi}) of {approximately} 4 {times} 10{sup 14} n/cm{sup 2} {center_dot} s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of {approximately} 6 s, but the requirement of immediate counting leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about {+-} 0.5 s) make irradiations of < 6 s less reliable. Therefore, the determination of these ultra-short-lived species in mixed matrices has not generally been made at HFIR. The authors have found that very short lived activation products can be produced easily during the period after reactor shutdown (SCRAM), but prior to the removal of spent fuel elements. During this 24- to 36-h period (dubbed the ``sleeping reactor``), neutrons are produced in the beryllium reflector by the reaction {sup 9}Be({gamma},n){sup 8}Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to {approximately} 1 {times} 10{sup 10} n/cm{sup 2} {center_dot} s within 1 h. By the time the fuel elements are removed, the flux has dropped to {approximately} 6 {times} 10{sup 8}. Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant.

  4. ISOTOPE CONVERSION DEVICE

    DOE Patents [OSTI]

    Wigner, E.P.; Young, G.J.; Ohlinger, L.A.

    1957-12-01

    This patent relates to nuclear reactors of tbe type utilizing a liquid fuel and designed to convert a non-thermally fissionable isotope to a thermally fissionable isotope by neutron absorption. A tank containing a reactive composition of a thermally fissionable isotope dispersed in a liquid moderator is disposed within an outer tank containing a slurry of a non-thermally fissionable isotope convertible to a thermally fissionable isotope by neutron absorption. A control rod is used to control the chain reaction in the reactive composition and means are provided for circulating and cooling the reactive composition and slurry in separate circuits.

  5. Control Means for Reactor

    DOE Patents [OSTI]

    Manley, J. H.

    1961-06-27

    An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

  6. Digital, remote control system for a 2-MW research reactor

    SciTech Connect (OSTI)

    Battle, R.E.; Corbett, G.K.

    1988-01-01

    A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs.

  7. POWER REACTOR

    DOE Patents [OSTI]

    Zinn, W.H.

    1958-07-01

    A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

  8. Fast Flux Test Facility (FFTF) Briefing Book 1 Summary

    SciTech Connect (OSTI)

    WJ Apley

    1997-12-01

    This report documents the results of evaluations preformed during 1997 to determine what, if an, future role the Fast Flux Test Facility (FFTF) might have in support of the Department of Energy’s tritium productions strategy. An evaluation was also conducted to assess the potential for the FFTF to produce medical isotopes. No safety, environmental, or technical issues associated with producing 1.5 kilograms of tritium per year in the FFTF have been identified that would change the previous evaluations by the Department of Energy, the JASON panel, or Putnam, Hayes & Bartlett. The FFTF can be refitted and restated by July 2002 for a total expenditure of $371 million, with an additional $64 million of startup expense necessary to incorporate the production of medical isotopes. Therapeutic and diagnostic applications of reactor-generated medical isotopes will increase dramatically over the next decade. Essential medical isotopes can be produced in the FFTF simultaneously with tritium production, and while a stand-alone medical isotope mission for the facility cannot be economically justified given current marker conditions, conservative estimates based on a report by Frost &Sullivan indicate that 60% of the annual operational costs (reactor and fuel supply) could be offset by revenues from medical isotope production within 10 yeas of restart. The recommendation of the report is for the Department of Energy to continue to maintain the FFTF in standby and proceed with preparation of appropriate Nations Environmental Policy Act documentation in full consultation with the public to consider the FFTF as an interim tritium production option (1.5 kilograms/year) with a secondary mission of producing medical isotopes.

  9. EIS-0310: Accomplishing Expanded Civilian Nuclear Energy Research and Development and Isotope Production Missions in the United States, Including the Role of the Fast Flux Test Facility

    Broader source: Energy.gov [DOE]

    This PEIS will evaluate the potential environmental impacts of the proposed enhancement of the existing infrastructure, including the possible role of the Fast Flux Test Facility (FFTF), located at...

  10. Method of fission product beta spectra measurements for predicting reactor anti-neutrino emission

    SciTech Connect (OSTI)

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

    2015-03-01

    The nuclear fission process that occurs in the core of nuclear reactors results in unstable, neutron-rich fission products that subsequently beta decay and emit electron antineutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to today's precision measurements of neutrino mixing angles. The prediction of the absolute flux and energy spectrum of the emitted reactor neutrinos hinges upon a series of seminal papers based on measurements performed in the 1970s and 1980s. The steadily improving reactor neutrino measurement techniques and recent reconsiderations of the agreement between the predicted and observed reactor neutrino flux motivates revisiting the underlying beta spectra measurements. A method is proposed to use an accelerator proton beam delivered to an engineered target to yield a neutron field tailored to reproduce the neutron energy spectrum present in the core of an operating nuclear reactor. Foils of the primary reactor fissionable isotopes placed in this tailored neutron flux will ultimately emit beta particles from the resultant fission products. Measurement of these beta particles in a time projection chamber with a perpendicular magnetic field provides a distinctive set of systematic considerations for comparison to the original seminal beta spectra measurements. Ancillary measurements such as gamma-ray emission and post-irradiation radiochemical analysis will further constrain the absolute normalization of beta emissions per fission. The requirements for unfolding the beta spectra measured with this method into a predicted reactor neutrino spectrum are explored.

  11. NEUTRON FLUX INTENSITY DETECTION

    DOE Patents [OSTI]

    Russell, J.T.

    1964-04-21

    A method of measuring the instantaneous intensity of neutron flux in the core of a nuclear reactor is described. A target gas capable of being transmuted by neutron bombardment to a product having a resonance absorption line nt a particular microwave frequency is passed through the core of the reactor. Frequency-modulated microwave energy is passed through the target gas and the attenuation of the energy due to the formation of the transmuted product is measured. (AEC)

  12. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    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.

  13. Neutronic Study of Slightly Modified Water Reactors and Application to Transition Scenarios

    SciTech Connect (OSTI)

    Chambon, Richard; Guillemin, Perrine; Nuttin, Alexis; Bidaud, A.

    2007-07-01

    In this paper we have studied slightly modified water reactors and their applications to transition scenarios. The PWR and CANDU reactors have been considered. New fuels based on Thorium have been tested: Thorium/Plutonium and Thorium/Uranium- 233, with different fissile isotope contents. Changes in the geometry of the assemblies were also explored to modify the moderation ratio, and consequently the neutron flux spectrum. A core equivalent assembly methodology was introduced as an exploratory approach and to reduce the computation time. Several basic safety analyses were also performed. We have finally developed a new scenario code, named OSCAR (Optimized Scenario Code for Advanced Reactors), to study the efficiency of these modified reactors in transition to Gen IV reactors or in symbiotic fleet. (authors)

  14. PRINCIPAL ISOTOPE SELECTION REPORT

    SciTech Connect (OSTI)

    K. D. Wright

    1998-08-28

    Utilizing nuclear fuel to produce power in commercial reactors results in the production of hundreds of fission product and transuranic isotopes in the spent nuclear fuel (SNF). When the SNF is disposed of in a repository, the criticality analyses could consider all of the isotopes, some principal isotopes affecting criticality, or none of the isotopes, other than the initial loading. The selected set of principal isotopes will be the ones used in criticality analyses of the SNF to evaluate the reactivity of the fuel/waste package composition and configuration. This technical document discusses the process used to select the principal isotopes and the possible affect that these isotopes could have on criticality in the SNF. The objective of this technical document is to discuss the process used to select the principal isotopes for disposal criticality evaluations with commercial SNF. The principal isotopes will be used as supporting information in the ''Disposal Criticality Analysis Methodology Topical Report'' which will be presented to the United States Nuclear Regulatory Commission (NRC) when approved by the United States Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM).

  15. (Reactor dosimetry)

    SciTech Connect (OSTI)

    West, C.D.

    1990-09-13

    The lead in most aspects of research reactor design and use passed from the USA about 15 years ago, soon after the construction of the HFIR and HFBR. The Europeans have consistently upgraded and improved their existing facilities and have built new ones including the HFR at Grenoble and ORPHEE at Saclay. They studied ultra-high flux concepts ({approximately}10{sup 20}/m{sup {minus}2}{center dot}s{sup {minus}1}) about 10 years ago, and are in the design phase of a new, highly efficient medium flux reactor to be built at Garching, near Munich in Germany. A visit was made to Interatom, the firm -- the equivalent of the Architect/Engineer for the ANS project -- responsible, under contract to the Technical University of Munich, for the new Munich reactor design. There are many similarities to the ANS design, and we reviewed and discussed technical and safety aspects of the two reactors. A request was made for some new, hitherto proprietary, experimental data on reactor thermal hydraulics and cooling that will be very valuable to the ANS project. I presented a seminar on the ANS project. A visit was made to Kernforschungszentrum Karlsruhe and knowledge was gained from Dr. Kuchle, a true pioneer of ultra-high flux reactor concepts, of their work. Dr. Kuchle kindly reviewed the ANS reference core and cooling system design (with favorable conclusions). I then talked with researchers working on materials irradiation damage and activation of structural materials by neutron irradiation, both key issues for the ANS. I was shown some new techniques they have developed for testing materials irradiation effects at high fluences, in a short time, using accelerated particle beams.

  16. Beta ray flux measuring device

    DOE Patents [OSTI]

    Impink, Jr., Albert J.; Goldstein, Norman P.

    1990-01-01

    A beta ray flux measuring device in an activated member in-core instrumentation system for pressurized water reactors. The device includes collector rings positioned about an axis in the reactor's pressure boundary. Activated members such as hydroballs are positioned within respective ones of the collector rings. A response characteristic such as the current from or charge on a collector ring indicates the beta ray flux from the corresponding hydroball and is therefore a measure of the relative nuclear power level in the region of the reactor core corresponding to the specific exposed hydroball within the collector ring.

  17. 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.

    2014-12-15

    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.

  18. B Reactor - Hanford Site

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

    Reactor About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs Contact Us 100 Area 118-K-1 Burial Ground 200 Area 222-S Laboratory 242-A Evaporator 300 Area 324 Building 325 Building 400 Area/Fast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim Storage Area Canyon Facilities Cold Test Facility D and DR Reactors Effluent Treatment Facility Environmental Restoration Disposal Facility F Reactor H

  19. Isotope Science

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

    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. Contact: Kevin John LANL Isotope Program Manager kjohn@lanl.gov 505-667-3602 Sponsored by the Department of Energy National Isotope Program http://www.nuclear.energy.gov/isotopes/nelsotopes2a.html Isotopes for Environmental Science Isotopes produced at Los Alamos National Laboratory are used as

  20. Isotopes Products

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

    Other isotopes that have recently shipped from LANL's isotope program include cadmium-109 (X-ray fluorescence sources), arsenic-72 (medical research), and sodium-22 (PET sources).

  1. Fast flux test facility radioisotope production and medical applications

    SciTech Connect (OSTI)

    Schenter, R.E.; Smith, S.G.; Tenforde, T.S.

    1997-12-01

    The Fast Flux Test Facility (FFTF) is a 400-MW, sodium-cooled reactor that operated successfully from 1982 to 1992, conducting work in support of the liquid-metal reactor industry by developing and testing fuel assemblies, control rods, and other core reactor components. Upon termination of this program, the primary mission of FFTF ended, and it was placed in a standby mode in 1993. However, in January 1997 the U.S. Secretary of Energy requested that FFTF be evaluated for a future mission that would consist of a primary goal of producing tritium for nuclear defense applications and a secondary goal of supplying medical isotopes for research and clinical applications. Production by FFTF of tritium for U.S. nuclear weapons would augment the dual-track strategy now under consideration for providing a long-term tritium supply in the United States (consisting of a light water reactor option and an accelerator option). A decision by the Secretary of Energy on proceeding with steps leading toward the possible reactivation of FFTF will be made before the end of 1998.

  2. THERMAL NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Spinrad, B.I.

    1960-01-12

    A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

  3. NUCLEAR REACTOR FUEL ELEMENT

    DOE Patents [OSTI]

    Currier, E.L. Jr.; Nicklas, J.H.

    1963-06-11

    A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

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

    SciTech Connect (OSTI)

    Ammoniums

    1999-06-01

    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

  5. Fission reactors and materials

    SciTech Connect (OSTI)

    Frost, B.R.T.

    1981-12-01

    The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions.

  6. D and DR Reactors - Hanford Site

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

    300 Area 324 Building 325 Building 400 AreaFast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim ...

  7. Radiation dosimetry at the BNL reactor facilities

    SciTech Connect (OSTI)

    Holden, N.E.; Hu, J.P.; Reciniello, R.N.; Greenberg, D.D.; Sengupta, S.; Farrell, K.; Greenwood, L.R.

    1999-07-01

    Neutron and gamma-ray dosimetry measurements have been performed at various facilities in the High Flux Beam Reactor (HFBR) and in the Brookhaven National Laboratory Medical Research Reactor (BMRR). These experimental results are discussed.

  8. 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.

    1982-03-31

    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.

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

    DOE Patents [OSTI]

    Caldwell, John T.; Kunz, Walter E.; Atencio, James D.

    1984-01-01

    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.

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

    SciTech Connect (OSTI)

    Weaver, Kevan Dean; Mac Donald, Philip Elsworth

    2002-10-01

    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.

  11. 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.

  12. STEAM FORMING NEUTRONIC REACTOR AND METHOD OF OPERATING IT

    DOE Patents [OSTI]

    Untermyer, S.

    1960-05-10

    The heterogeneous reactor is liquid moderated and cooled by a steam forming coolant and is designed to produce steam from the coolant directly within the active portion of the reactor while avoiding the formation of bubbles in the liquid moderator. This reactor achieves inherent stability as a result of increased neutron leakage and increased neutron resonance absorption in the U/sup 238/ fuel with the formation of bubbles. The invention produces certain conditions under which the formation of vapor bubbles as a result of a neutron flux excursion from the injection of a reactivity increment into the reactor will operate to nullify the reactivity increment within a sufficiently short period of time to prevent unsafe reactor operating conditions from developing. This is obtained by disposing a plurality of fuel elements within a mass of steam forming coolant in the core with the ratio of the volume of steam forming coolant to the volume of fissionable isotopes being within the range yielding a multiplication factor greater than unity and a negative reactivity to core void coefficient at the boiling temperature of the coolant.

  13. Transportation of medical isotopes

    SciTech Connect (OSTI)

    Nielsen, D.L.

    1997-11-19

    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. Nuclear reactor shutdown system

    DOE Patents [OSTI]

    Bhate, Suresh K.; Cooper, Martin H.; Riffe, Delmar R.; Kinney, Calvin L.

    1981-01-01

    An inherent shutdown system for a nuclear reactor having neutron absorbing rods affixed to an armature which is held in an upper position by a magnetic flux flowing through a Curie temperature material. The Curie temperature material is fixedly positioned about the exterior of an inner duct in an annular region through which reactor coolant flows. Elongated fuel rods extending from within the core upwardly toward the Curie temperature material are preferably disposed within the annular region. Upon abnormal conditions which result in high neutron flux and coolant temperature, the Curie material loses its magnetic permeability, breaking the magnetic flux path and allowing the armature and absorber rods to drop into the core, thus shutting down the fissioning reaction. The armature and absorber rods are retrieved by lowering the housing for the electromagnet forming coils which create a magnetic flux path which includes the inner duct wall. The coil housing then is raised, resetting the armature.

  15. High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

    SciTech Connect (OSTI)

    Guillen, Donna; Greenwood, Lawrence R.; Parry, James

    2014-06-22

    A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated for up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.

  16. Isotope separation

    DOE Patents [OSTI]

    Bartlett, Rodney J.; Morrey, John R.

    1978-01-01

    A method and apparatus is described for separating gas molecules containing one isotope of an element from gas molecules containing other isotopes of the same element in which all of the molecules of the gas are at the same electronic state in their ground state. Gas molecules in a gas stream containing one of the isotopes are selectively excited to a different electronic state while leaving the other gas molecules in their original ground state. Gas molecules containing one of the isotopes are then deflected from the other gas molecules in the stream and thus physically separated.

  17. Accurate determination of Curium and Californium isotopic ratios by inductively coupled plasma quadrupole mass spectrometry (ICP-QMS) in 248Cm samples for transmutation studies

    SciTech Connect (OSTI)

    Gourgiotis, A.; Isnard, H.; Aubert, M.; Dupont, E.; AlMahamid, I.; Cassette, P.; Panebianco, S.; Letourneau, A.; Chartier, F.; Tian, G.; Rao, L.; Lukens, W.

    2011-02-01

    The French Atomic Energy Commission has carried out several experiments including the mini-INCA (INcineration of Actinides) project for the study of minor-actinide transmutation processes in high intensity thermal neutron fluxes, in view of proposing solutions to reduce the radiotoxicity of long-lived nuclear wastes. In this context, a Cm sample enriched in {sup 248}Cm ({approx}97 %) was irradiated in thermal neutron flux at the High Flux Reactor (HFR) of the Laue-Langevin Institute (ILL). This work describes a quadrupole ICP-MS (ICP-QMS) analytical procedure for precise and accurate isotopic composition determination of Cm before sample irradiation and of Cm and Cf after sample irradiation. The factors that affect the accuracy and reproducibility of isotopic ratio measurements by ICP-QMS, such as peak centre correction, detector dead time, mass bias, abundance sensitivity and hydrides formation, instrumental background, and memory blank were carefully evaluated and corrected. Uncertainties of the isotopic ratios, taking into account internal precision of isotope ratio measurements, peak tailing, and hydrides formations ranged from 0.3% to 1.3%. This uncertainties range is quite acceptable for the nuclear data to be used in transmutation studies.

  18. Isotope geochemistry

    SciTech Connect (OSTI)

    Cole, D.R.; Curtis, D.B.; DePaolo, D.J.; Gerlach, T.M.; Laul, J.C.; Shaw, H.; Smith, B.M.; Sturchio, N.C.

    1990-09-01

    This document represents the consensus of members of the ad hoc Committee on Isotope Geochemistry in the US Department of Energy; the committee is composed of researchers in isotope geochemistry from seven of the national laboratories. Information included in this document was presented at workshops at Lawrence Berkeley Laboratory (April 1989) and at Los Alamos National Laboratory (August 1989).

  19. Peaceful Uses of the Atom and Atoms for Peace

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

    Production of Medical Radioisotopes in the ORNL High Flux Isotope Reactor (HFIR) for Cancer Treatment and Arterial Restenosis Therapy after PTCA The High Flux Isotope Reactor ...

  20. Chromatographic hydrogen isotope separation

    DOE Patents [OSTI]

    Aldridge, F.T.

    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 column. The addition of an inert metal to dilute the hydride improves performance of the column. A large scale multi-stage chromatographic separation process run as a secondary process off a hydrogen feedstream from an industrial plant which uses large volumes of hydrogen cn produce large quantities of heavy water at an effective cost for use in heavy water reactors.

  1. Chromatographic hydrogen isotope separation

    DOE Patents [OSTI]

    Aldridge, Frederick T.

    1981-01-01

    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.

  2. Laser separation of medical isotopes

    SciTech Connect (OSTI)

    Eerkens, J.W.; Puglishi, D.A.; Miller, W.H.

    1996-12-31

    There is an increasing demand for different separated isotopes as feed material for reactor and cyclotron-produced radioisotopes used by a fast-growing radiopharmaceutical industry. One new technology that may meet future demands for medical isotopes is molecular laser isotope separation (MLIS). This method was investigated for the enrichment of uranium in the 1970`s and 1980s by Los Alamos National Laboratory, Isotope Technologies, and others around the world. While South Africa and Japan have continued the development of MLIS for uranium and are testing pilot units, around 1985 the United States dropped the LANL MLIS program in favor of AVLIS (atomic vapor LIS), which uses electron-beam-heated uranium metal vapor. AVLIS appears difficult and expensive to apply to most isotopes of medical interest, however, whereas MLIS technology, which is based on cooled hexafluorides or other gaseous molecules, can be adapted more readily. The attraction of MLIS for radiopharmaceutical firms is that it allows them to operate their own dedicated separators for small-quantity productions of critical medical isotopes, rather than having to depend on large enrichment complexes run by governments, which are only optimal for large-quantity productions. At the University of Missouri, the authors are investigating LIS of molybdenum isotopes using MoF{sub 6}, which behaves in a way similar to UF{sub 6}, studied in the past.

  3. Isotopes Products

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

    Isotopes Products Isotopes Products Isotopes produced at Los Alamos National Laboratory are saving lives, advancing cutting-edge research and keeping the U.S. safe. Products stress and rest Stress and rest Rb-82 PET images in a patient with dipyridamole stress-inducible lateral wall and apical ischemia. (http://www.fac.org.ar/scvc/llave/image/machac/machaci.htm#f2,3,4) Strontium-82 is supplied to our customers for use in Sr-82/Rb-82 generator technologies. The generators in turn are supplied to

  4. Nuclear Archeology for CANDU Power Reactors

    SciTech Connect (OSTI)

    Broadhead, Bryan L

    2011-01-01

    The goal of this work is the development of so-called 'nuclear archeology' techniques to predict the irradiation history of both fuel-related and non-fuel-related materials irradiated in the CANDU (CANada Deuterium Uranium) family of nuclear reactors. In this application to CANDU-type reactors, two different scenarios for the collection of the appropriate data for use in these procedures will be assumed: the first scenario is the removal of the pressure tubes, calandria tubes, or fuel cladding and destructive analysis of the activation products contained in these structural materials; the second scenario is the nondestructive analysis (NDA) of the same hardware items via high-resolution gamma ray scans. There are obvious advantages and disadvantages for each approach; however, the NDA approach is the central focus of this work because of its simplicity and lack of invasiveness. The use of these techniques along with a previously developed inverse capability is expected to allow for the prediction of average flux levels and irradiation time, and the total fluence for samples where the values of selected isotopes can be measured.

  5. Apparatus for isotopic alteration of mercury vapor

    DOE Patents [OSTI]

    Grossman, Mark W.; George, William A.; Marcucci, Rudolph V.

    1988-01-01

    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.

  6. K-East and K-West Reactors - Hanford Site

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

    300 Area 324 Building 325 Building 400 AreaFast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim ...

  7. Laser Isotope Enrichment for Medical and Industrial Applications

    SciTech Connect (OSTI)

    Leonard Bond

    2006-07-01

    Laser Isotope Enrichment for Medical and Industrial Applications by Jeff Eerkens (University of Missouri), Jay Kunze (Idaho State University), and Leonard Bond (Idaho National Laboratory) The principal isotope enrichment business in the world is the enrichment of uranium for commercial power reactor fuels. However, there are a number of other needs for separated isotopes. Some examples are: 1) Pure isotopic targets for irradiation to produce medical radioisotopes. 2) Pure isotopes for semiconductors. 3) Low neutron capture isotopes for various uses in nuclear reactors. 4) Isotopes for industrial tracer/identification applications. Examples of interest to medicine are targets to produce radio-isotopes such as S-33, Mo-98, Mo-100, W-186, Sn-112; while for MRI diagnostics, the non-radioactive Xe-129 isotope is wanted. For super-semiconductor applications some desired industrial isotopes are Si-28, Ga-69, Ge-74, Se-80, Te-128, etc. An example of a low cross section isotope for use in reactors is Zn-68 as a corrosion inhibitor material in nuclear reactor primary systems. Neutron activation of Ar isotopes is of interest in industrial tracer and diagnostic applications (e.g. oil-logging). . In the past few years there has been a sufficient supply of isotopes in common demand, because of huge Russian stockpiles produced with old electromagnetic and centrifuge separators previously used for uranium enrichment. Production of specialized isotopes in the USA has been largely accomplished using old ”calutrons” (electromagnetic separators) at Oak Ridge National Laboratory. These methods of separating isotopes are rather energy inefficient. Use of lasers for isotope separation has been considered for many decades. None of the proposed methods have attained sufficient proof of principal status to be economically attractive to pursue commercially. Some of the authors have succeeded in separating sulfur isotopes using a rather new and different method, known as condensation

  8. Plug Flow Reactor Simulator

    Energy Science and Technology Software Center (OSTI)

    1996-07-30

    PLUG is a computer program that solves the coupled steady state continuity, momentum, energy, and species balance equations for a plug flow reactor. Both homogeneous (gas-phase) and heterogenous (surface) reactions can be accommodated. The reactor may be either isothermal or adiabatic or may have a specified axial temperature or heat flux profile; alternatively, an ambient temperature and an overall heat-transfer coefficient can be specified. The crosssectional area and surface area may vary with axial position,more » and viscous drag is included. Ideal gas behavior and surface site conservation are assumed.« less

  9. Enforcement Letter, International Isotopes Idaho Inc- August 20, 1999

    Broader source: Energy.gov [DOE]

    Issued to International Isotopes Idaho, Inc. related to the Relocation of an Irradiated Pellet at the Test Reactor Area Hot Cell Facility at the Idaho National Engineering and Environmental Laboratory

  10. ISOTOPE SEPARATORS

    DOE Patents [OSTI]

    Bacon, C.G.

    1958-08-26

    An improvement is presented in the structure of an isotope separation apparatus and, in particular, is concerned with a magnetically operated shutter associated with a window which is provided for the purpose of enabling the operator to view the processes going on within the interior of the apparatus. The shutier is mounted to close under the force of gravity in the absence of any other force. By closing an electrical circuit to a coil mouated on the shutter the magnetic field of the isotope separating apparatus coacts with the magnetic field of the coil to force the shutter to the open position.

  11. I ISOTOPES

    Office of Legacy Management (LM)

    fl6-6 ' , WTELEEYNE I ISOTOPES i - ' 50<77 /,' y. 6 IWL-5025-473 SUBSURFACE URASIUM OJ: THE GROUNDS OF NL BEARINGS, ALBAh'Y Heyitt Iv. Jeter Douglas M. Eagleson Fred J. Frullo TELEDYNE ISOTOPES 50 VAK BUREN A\!EMJE WESTKOOD, NEK JERSEY 07675 7 Dcccmhcr 1953 Prepnrcd for NL f%carings/NL Tndustrics, Inc. 1130 CCVltrill AXr~lMIC Allmy, New York 12205 TABLE OF CONTEhTS 1.0 INTRODUCTION 2.0 METHODS 2.1 Soil Sampling 2.2 Sample Preparation 2.3 Analysis of Samples 3.0 RESULTS 4.0 SUMMARY REFERENCES

  12. Evaluation of in-vessel corium retention through external reactor vessel cooling for integral reactor

    SciTech Connect (OSTI)

    Park, R. J.; Lee, J. R.; Kim, S. B.; Jin, Y.; Kim, H. Y.

    2012-07-01

    In-vessel corium retention through external reactor vessel cooling (IVR-ERVC) for a small integral reactor has been evaluated to determine the thermal margin for the prevention of a reactor vessel failure. A thermal load analysis from the corium pool to the outer reactor vessel wall in the lower plenum of the reactor vessel has been performed to determine the heat flux distribution. The critical heat flux (CHF) on the outer reactor vessel wall has been determined to fix the maximum heat removal rate through the external coolant between the outer reactor vessel and the insulation of the reactor vessel. Finally, the thermal margin has been evaluated by comparison of the thermal load with the maximum heat removal rate of the CHF on the outer reactor vessel wall. The maximum heat flux from the corium pool to the outer reactor vessel is estimated at approximately 0.25 MW/m{sup 2} in the metallic layer because of the focusing effect. The CHF of the outer reactor vessel is approximately 1.1 MW/m{sup 2} because of a two phase natural circulation mass flow. Since the thermal margin for the IVR-ERVC is sufficient, the reactor vessel integrity is maintained during a severe accident of a small integral reactor. (authors)

  13. GUM Analysis for TIMS and SIMS Isotopic Ratios in Graphite

    SciTech Connect (OSTI)

    Heasler, Patrick G.; Gerlach, David C.; Cliff, John B.; Petersen, Steven L.

    2007-04-01

    This report describes GUM calculations for TIMS and SIMS isotopic ratio measurements of reactor graphite samples. These isotopic ratios are used to estimate reactor burn-up, and currently consist of various ratios of U, Pu, and Boron impurities in the graphite samples. The GUM calculation is a propagation of error methodology that assigns uncertainties (in the form of standard error and confidence bound) to the final estimates.

  14. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  15. Selected Isotopes for Optimized Fuel Assembly Tags

    SciTech Connect (OSTI)

    Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

    2008-10-01

    In support of our ongoing signatures project we present information on 3 isotopes selected for possible application in optimized tags that could be applied to fuel assemblies to provide an objective measure of burnup. 1. Important factors for an optimized tag are compatibility with the reactor environment (corrosion resistance), low radioactive activation, at least 2 stable isotopes, moderate neutron absorption cross-section, which gives significant changes in isotope ratios over typical fuel assembly irradiation levels, and ease of measurement in the SIMS machine 2. From the candidate isotopes presented in the 3rd FY 08 Quarterly Report, the most promising appear to be Titanium, Hafnium, and Platinum. The other candidate isotopes (Iron, Tungsten, exhibited inadequate corrosion resistance and/or had neutron capture cross-sections either too high or too low for the burnup range of interest.

  16. 400 Area/Fast Flux Test Facility - Hanford Site

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

    325 Building 400 AreaFast Flux Test Facility 618-10 ... Test Facility D and DR Reactors Effluent Treatment ... (thermal) liquid-metal (sodium)-cooled nuclear research ...

  17. --No Title--

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

    29 Top of the HFIR reactor. Aerial view of the ATRC reactor core and bridge. Oak Ridge National Laboratory High Flux Isotope Reactor High Flux Isotope Reactor HFIR is a versatile ...

  18. APPARATUS FOR CONTROLLING NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Dietrich, J.R.; Harrer, J.M.

    1958-09-16

    A device is described for rapidly cortrolling the reactivity of an active portion of a reactor. The inveniion consists of coaxially disposed members each having circumferenital sections of material having dlfferent neutron absorbing characteristics and means fur moving the members rotatably and translatably relative to each other within the active portion to vary the neutron flux therein. The angular and translational movements of any member change the neutron flux shadowing effect of that member upon the other member.

  19. Independent Oversight Review, Oak Ridge National Laboratory- January 2013

    Broader source: Energy.gov [DOE]

    Review of the Oak Ridge National Laboratory High Flux Isotope Reactor Implementation Verification Review Processes

  20. NUCLEAR REACTOR CONTROL SYSTEM

    DOE Patents [OSTI]

    Howard, D.F.; Motta, E.E.

    1961-06-27

    A method for controlling the excess reactivity in a nuclear reactor throughout the core life while maintaining the neutron flux distribution at the desired level is described. The control unit embodies a container having two electrodes of different surface area immersed in an electrolytic solution of a good neutron sbsorbing metal ion such as boron, gadolinium, or cadmium. Initially, the neutron absorber is plated on the larger electrode to control the greater neutron flux of a freshly refueled core. As the fuel burns up, the excess reactivity decreases and the neutron absorber is then plated onto the smaller electrode so that the number of neutrons absorbed also decreases. The excess reactivity in the core may thus be maintained without the introduction of serious perturbations in the neutron flux distributibn.

  1. Fast flux locked loop

    DOE Patents [OSTI]

    Ganther, Jr., Kenneth R.; Snapp, Lowell D.

    2002-09-10

    A flux locked loop for providing an electrical feedback signal, the flux locked loop employing radio-frequency components and technology to extend the flux modulation frequency and tracking loop bandwidth. The flux locked loop of the present invention has particularly useful application in read-out electronics for DC SQUID magnetic measurement systems, in which case the electrical signal output by the flux locked loop represents an unknown magnetic flux applied to the DC SQUID.

  2. Knowledge Preservation at the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Wootan, David W.; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.; Nelson, Joseph V.; Polzin, David L.

    2012-01-30

    The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to operate in the United States, from 1982 to 1992. The technologies employed in designing and constructing this reactor, along with information obtained from tests conducted during its operation, are currently being secured and archived by the Department of Energy's Office of Nuclear Energy. This report is one in a series documenting the overall project efforts to retrieve and preserve critical information related to advanced reactors

  3. Knowledge Preservation at the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Wootan, David W.; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.; Nelson, Joseph V.; Polzin, David L.

    2011-11-30

    The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to operate in the United States, from 1982 to 1992. The technologies employed in designing and constructing this reactor, along with information obtained from tests conducted during its operation, are currently being secured and archived by the Department of Energy's Office of Nuclear Energy. This report provides a status update documenting the overall project efforts to retrieve and preserve critical information related to advanced reactors.

  4. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    DOE Patents [OSTI]

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  5. Reactor-Produced Medical Radionuclides

    SciTech Connect (OSTI)

    Mirzadeh, Saed; Mausner, Leonard; Garland, Marc A

    2011-01-01

    The therapeutic use of radionuclides in nuclear medicine, oncology and cardiology is the most rapidly growing use of medical radionuclides. Since most therapeutic radionuclides are neutron rich and decay by beta emission, they are reactor-produced. This chapter deals mainly with production approaches with neutrons. Neutron interactions with matter, neutron transmission and activation rates, and neutron spectra of nuclear reactors are discussed in some detail. Further, a short discussion of the neutron-energy dependence of cross sections, reaction rates in thermal reactors, cross section measurements and flux monitoring, and general equations governing the reactor production of radionuclides are presented. Finally, the chapter is concluded by providing a number of examples encompassing the various possible reaction routes for production of a number of medical radionuclides in a reactor.

  6. CONTROL SYSTEM FOR NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Crever, F.E.

    1962-05-01

    BS>A slow-acting shim rod for control of major variations in reactor neutron flux and a fast-acting control rod to correct minor flux variations are employed to provide a sensitive, accurate control system. The fast-acting rod is responsive to an error signal which is produced by changes in the neutron flux from a predetermined optimum level. When the fast rod is thus actuated in a given direction, means is provided to actuate the slow-moving rod in that direction to return the fast rod to a position near the midpoint of its control range. (AEC)

  7. BOILING REACTORS

    DOE Patents [OSTI]

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  8. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

  9. Nuclear reactor shield including magnesium oxide

    DOE Patents [OSTI]

    Rouse, Carl A.; Simnad, Massoud T.

    1981-01-01

    An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.

  10. Energy deposition in STARFIRE reactor components

    SciTech Connect (OSTI)

    Gohar, Y.; Brooks, J.N.

    1985-04-01

    The energy deposition in the STARFIRE commercial tokamak reactor was calculated based on detailed models for the different reactor components. The heat deposition and the 14 MeV neutron flux poloidal distributions in the first wall were obtained. The poloidal surface heat load distribution in the first wall was calculated from the plasma radiation. The Monte Carlo method was used for the calculation to allow an accurate modeling for the reactor geometry.

  11. CONVECTION REACTOR

    DOE Patents [OSTI]

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

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

    SciTech Connect (OSTI)

    Spencer, Khalil J.; Rim, Jung Ho; Porterfield, Donivan R.; Roback, Robert Clifford; Boukhalfa, Hakim; Stanley, Floyd E.

    2015-06-29

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

  13. Pulse flux measuring device

    DOE Patents [OSTI]

    Riggan, William C.

    1985-01-01

    A device for measuring particle flux comprises first and second photodiode detectors for receiving flux from a source and first and second outputs for producing first and second signals representing the flux incident to the detectors. The device is capable of reducing the first output signal by a portion of the second output signal, thereby enhancing the accuracy of the device. Devices in accordance with the invention may measure distinct components of flux from a single source or fluxes from several sources.

  14. Predicting Activation of Experiments Inside the Annular Core Research Reactor

    SciTech Connect (OSTI)

    Greenberg, Joseph Isaac

    2015-11-01

    The objective of this thesis is to create a program to quickly estimate the radioactivity and decay of experiments conducted inside of the Annular Core Research Reactor at Sandia National Laboratories and eliminate the need for users to write code. This is achieved by model the neutron fluxes in the reactor’s central cavity where experiments are conducted for 4 different neutron spectra using MCNP. The desired neutron spectrum, experiment material composition, and reactor power level are then input into CINDER2008 burnup code to obtain activation and decay information for every isotope generated. DREAD creates all of the files required for CINDER2008 through user selected inputs in a graphical user interface and executes the program for the user and displays the resulting estimation for dose rate at various distances. The DREAD program was validated by weighing and measuring various experiments in the different spectra and then collecting dose rate information after they were irradiated and comparing it to the dose rates that DREAD predicted. The program provides results with an average of 17% higher estimates than the actual values and takes seconds to execute.

  15. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fraas, A.P.; Mills, C.B.

    1961-11-21

    A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

  16. REACTOR COOLING

    DOE Patents [OSTI]

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  17. ENGINEERING TEST REACTOR

    DOE Patents [OSTI]

    De Boisblanc, D.R.; Thomas, M.E.; Jones, R.M.; Hanson, G.H.

    1958-10-21

    Heterogeneous reactors of the type which is both cooled and moderated by the same fluid, preferably water, and employs highly enriched fuel are reported. In this design, an inner pressure vessel is located within a main outer pressure vessel. The reactor core and its surrounding reflector are disposed in the inner pressure vessel which in turn is surrounded by a thermal shield, Coolant fluid enters the main pressure vessel, fiows downward into the inner vessel where it passes through the core containing tbe fissionable fuel assemblies and control rods, through the reflector, thence out through the bottom of the inner vessel and up past the thermal shield to the discharge port in the main vessel. The fuel assemblles are arranged in the core in the form of a cross having an opening extending therethrough to serve as a high fast flux test facility.

  18. WATER BOILER REACTOR

    DOE Patents [OSTI]

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  19. Method for separating isotopes

    DOE Patents [OSTI]

    Jepson, B.E.

    1975-10-21

    Isotopes are separated by contacting a feed solution containing the isotopes with a cyclic polyether wherein a complex of one isotope is formed with the cyclic polyether, the cyclic polyether complex is extracted from the feed solution, and the isotope is thereafter separated from the cyclic polyether.

  20. Stable isotope studies

    SciTech Connect (OSTI)

    Ishida, T.

    1992-01-01

    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.

  1. Neutrino-4 experiment on the search for a sterile neutrino at the SM-3 reactor

    SciTech Connect (OSTI)

    Serebrov, A. P. Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Chernyi, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Tsinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; and others

    2015-10-15

    In view of the possibility of the existence of a sterile neutrino, test measurements of the dependence of the reactor antineutrino flux on the distance from the reactor core has been performed on SM-2 reactor with the Neutrino-2 detector model in the range of 6–11 m. Prospects of the search for reactor antineutrinos at short distances have been discussed.

  2. Fast Reactor Technology Preservation

    SciTech Connect (OSTI)

    Wootan, David W.; Omberg, Ronald P.

    2008-01-11

    There is renewed worldwide interest in developing and implementing a new generation of advanced fast reactors. International cooperative efforts are underway such as the Global Nuclear Energy Partnership (GNEP). Advanced computer modeling and simulation efforts are a key part of these programs. A recognized and validated set of Benchmark Cases are an essential component of such modeling efforts. Testing documentation developed during the operation of the Fast Flux Test Facility (FFTF) provide the information necessary to develop a very useful set of Benchmark Cases.

  3. Physics of sup 238 Pu production in the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Rawlins, J.A.; Schmittroth, F.A.; Mann, F.M.; Schenter, R.E.; Lu, A.H.; Carter, L.L.; Wootan, D.W.; Schwarz, R.A.; Brager, H.R.; Matsumoto, W.Y.

    1989-11-01

    Westinghouse Hanford Company conducted an assessment of producing {sup 238}Pu in the Fast Flux Test Facility (FFTF), a U.S. Department of Energy (DOE) reactor located near Richland, Washington. The goal of the assessment was to determine whether the FFTF can produce at least 15 kg/yr of {sup 238}Pu to support the needs of the U.S. space program. Plutonium-238, with its 87.7-yr half-life and relatively pure alpha-particle decay mode, is an ideal power source for deep-space missions. The DOE is evaluating options for future {sup 238}Pu production, and the FFTF is a preferred candidate. The key technical issue for FFTF production is the isotopic purity of the product plutonium. FFTF production of at least 15 kg/yr of {sup 238}Pu is feasible. An FFTF physics test was completed and will reduce the large calculational uncertainties in {sup 236}Pu content, and the final test results will allow final production assembly design optimization. Use of the FFTF for {sup 238}Pu production can satisfy the needs of the U.S. space program for many years with a modern reactor that has an outstanding operational record.

  4. Isotope separation by photochromatography

    DOE Patents [OSTI]

    Suslick, K.S.

    1975-10-03

    A photochromatographic method for isotope separation is described. An isotopically mixed molecular species is adsorbed on an adsorptive surface, and the adsorbed molecules are irradiated with radiation of a predetermined wavelength which will selectively excite desired isotopic species. Sufficient energy is transferred to the excited molecules to desorb them from the surface and thus separate them from the undesired isotopic species. The method is particularly applicable to the separation of hydrogen isotopes. (BLM)

  5. Isotope separation by photochromatography

    DOE Patents [OSTI]

    Suslick, Kenneth S.

    1977-01-01

    An isotope separation method which comprises physically adsorbing an isotopically mixed molecular species on an adsorptive surface and irradiating the adsorbed molecules with radiation of a predetermined wavelength which will selectively excite a desired isotopic species. Sufficient energy is transferred to the excited molecules to desorb them from the surface and thereby separate them from the unexcited undesired isotopic species. The method is particularly applicable to the separation of hydrogen isotopes.

  6. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    SciTech Connect (OSTI)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  7. Laser-assisted isotope separation of tritium

    DOE Patents [OSTI]

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

    1983-01-01

    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.

  8. CONTROL MEANS FOR A NUCLEAR REACTOR

    DOE Patents [OSTI]

    Teitel, R.J.

    1961-09-01

    A control means is described for a reactor which employs a liquid fuel consisting of a fissile isotope in a liquid bismuth solvent. The liquid fuel is contained in a plurality of tubular vessels. Control is effected by inserting plungers in the vessels to displace the liquid fuel and provide a critical or non- critical fuel configuration as desired.

  9. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  10. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  11. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Wigner, E.P.

    1960-11-22

    A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

  12. REACTOR SHIELD

    DOE Patents [OSTI]

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  13. Reactor apparatus

    DOE Patents [OSTI]

    Echtler, J. Paul

    1981-01-01

    A reactor apparatus for hydrocracking a polynuclear aromatic hydrocarbonaceous feedstock to produce lighter hydrocarbon fuels by contacting the hydrocarbonaceous feedstock with hydrogen in the presence of a molten metal halide catalyst.

  14. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  15. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  16. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  17. Preliminary Notice of Violation, International Isotopes Idaho, Inc.- EA-2000-04

    Broader source: Energy.gov [DOE]

    Issued to International Isotopes Idaho, Inc., related 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 Engineering and Environmental Laboratory, May 19, 2000

  18. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  19. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  20. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

  1. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  2. ARM - Measurement - Methane flux

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

    flux ARM Data Discovery Browse Data Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Methane flux Vertical flux of methane near the surface due to turbulent transport. Categories Surface Properties, Atmospheric Carbon Instruments The above measurement is considered scientifically relevant for the following instruments. Refer to the datastream (netcdf) file headers of each instrument for a list of all available measurements, including

  3. Startup Testing of the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.; Polzin, David L.

    2010-06-30

    This paper is one in a series documenting the current effort to retrieve, secure, and preserve critical information related to advanced reactors. . Information from this testing is being retrieved under the Fuel Cycle Research and Development (FCRD) program conducted by the Office of Nuclear Energy (NE) of the DOE. The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE).

  4. Manus Water Isotope Investigation

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

    9 Manus Water Isotope Investigation Field Campaign Report JL Conroy D Noone KM Cobb March ... DOESC-ARM-15-079 Manus Water Isotope Investigation Field Campaign Report JL Conroy, ...

  5. Manus Water Isotope Investigation

    Office of Scientific and Technical Information (OSTI)

    ENERGY Office of Science DOESC-ARM-15-079 Manus Water Isotope Investigation Field ... DOESC-ARM-15-079 Manus Water Isotope Investigation Field Campaign Report JL Conroy, ...

  6. Review of the Oak Ridge National Laboratory High Flux Isotope...

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

    for implementation of new or revised safety basis, including the performance of a graded IVR. The procedure also requires managers to support scheduled line management and...

  7. Recovery of Information from the Fast Flux Test Facility for the Advanced Fuel Cycle Initiative

    SciTech Connect (OSTI)

    Nielsen, Deborah L.; Makenas, Bruce J.; Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.

    2009-09-30

    The Fast Flux Test Facility is the most recent Liquid Metal Reactor to operate in the United States. Information from the design, construction, and operation of this reactor was at risk as the facilities associated with the reactor are being shut down. The Advanced Fuel Cycle Initiative is a program managed by the Office of Nuclear Energy of the U.S. Department of Energy with a mission to develop new fuel cycle technologies to support both current and advanced reactors. Securing and preserving the knowledge gained from operation and testing in the Fast Flux Test Facility is an important part of the Knowledge Preservation activity in this program.

  8. Thermionic switched self-actuating reactor shutdown system

    DOE Patents [OSTI]

    Barrus, Donald M. (San Jose, CA); Shires, Charles D. (San Jose, CA); Brummond, William A. (Livermore, CA)

    1989-01-01

    A self-actuating reactor shutdown system incorporating a thermionic switched electromagnetic latch arrangement which is responsive to reactor neutron flux changes and to reactor coolant temperature changes. The system is self-actuating in that the sensing thermionic device acts directly to release (scram) the control rod (absorber) without reference or signal from the main reactor plant protective and control systems. To be responsive to both temperature and neutron flux effects, two detectors are used, one responsive to reactor coolant temperatures, and the other responsive to reactor neutron flux increase. The detectors are incorporated into a thermionic diode connected electrically with an electromagnetic mechanism which under normal reactor operating conditions holds the the control rod in its ready position (exterior of the reactor core). Upon reaching either a specified temperature or neutron flux, the thermionic diode functions to short-circuit the electromagnetic mechanism causing same to lose its holding power and release the control rod, which drops into the reactor core region under gravitational force.

  9. Level 1 transient model for a molybdenum-99 producing aqueous homogeneous reactor and its applicability to the tracy reactor

    SciTech Connect (OSTI)

    Nygaard, E. T.; Williams, M. M. R.; Angelo, P. L.

    2012-07-01

    Babcock and Wilcox Technical Services Group (B and W) has identified aqueous homogeneous reactors (AHRs) as a technology well suited to produce the medical isotope molybdenum 99 (Mo-99). AHRs have never been specifically designed or built for this specialized purpose. However, AHRs have a proven history of being safe research reactors. In fact, in 1958, AHRs had 'a longer history of operation than any other type of research reactor using enriched fuel' and had 'experimentally demonstrated to be among the safest of all various type of research reactor now in use [1].' A 'Level 1' model representing B and W's proposed Medical Isotope Production System (MIPS) reactor has been developed. The Level 1 model couples a series of differential equations representing neutronics, temperature, and voiding. Neutronics are represented by point reactor kinetics while temperature and voiding terms are axially varying (one-dimensional). While this model was developed specifically for the MIPS reactor, its applicability to the Japanese TRACY reactor was assessed. The results from the Level 1 model were in good agreement with TRACY experimental data and found to be conservative over most of the time domains considered. The Level 1 model was used to study the MIPS reactor. An analysis showed the Level 1 model agreed well with a more complex computational model of the MIPS reactor (a FETCH model). Finally, a significant reactivity insertion was simulated with the Level 1 model to study the MIPS reactor's time-dependent response. (authors)

  10. GUM Analysis for SIMS Isotopic Ratios in BEP0 Graphite Qualification Samples, Round 2

    SciTech Connect (OSTI)

    Gerlach, David C.; Heasler, Patrick G.; Reid, Bruce D.

    2009-01-01

    This report describes GUM calculations for TIMS and SIMS isotopic ratio measurements of reactor graphite samples. These isotopic ratios are used to estimate reactor burn-up, and currently consist of various ratios of U, Pu, and Boron impurities in the graphite samples. The GUM calculation is a propagation of error methodology that assigns uncertainties (in the form of standard error and confidence bound) to the final estimates.

  11. FUEL CYCLE ISOTOPE EVOLUTION BY TRANSMUTATION DYNAMICS OVER MULTIPLE RECYCLES

    SciTech Connect (OSTI)

    Samuel Bays; Steven Piet; Amaury Dumontier

    2010-06-01

    Because all actinides have the ability to fission appreciably in a fast neutron spectrum, these types of reactor systems are usually not associated with the buildup of higher mass actinides: curium, berkelium and californium. These higher actinides have high specific decay heat power, gamma and neutron source strengths, and are usually considered as a complication to the fuel manufacturing and transportation of fresh recycled transuranic fuel. This buildup issue has been studied widely for thermal reactor fuels. However, recent studies have shown that the transmutation physics associated with "gateway isotopes" dictates Cm-Bk-Cf buildup, even in fast burner reactors. Assuming a symbiotic fuel relationship with light water reactors (LWR), Pu-242 and Am-243 are formed in the LWRs and then are externally fed to the fast reactor as part of its overall transuranic fuel supply. These isotopes are created much more readily in a thermal than in fast spectrum systems due to the differences in the fast fission (i.e., above the fission threshold for non-fissile actinides) contribution. In a strictly breeding fast reactor this dependency on LWR transuranics would not exist, and thus avoids the introduction of LWR derived gateway isotopes into the fast reactor system. However in a transuranic burning fast reactor, the external supply of these gateway isotopes behaves as an external driving force towards the creation and build-up of Cm-Bk-Cf in the fuel cycle. It was found that though the Cm-Bk-Cf concentration in the equilibrium fuel cycle is dictated by the fast neutron spectrum, the time required to reach that equilibrium concentration is dictated by recycle, transmutation and decay storage dynamics.

  12. Advanced Reactors Transition Program Resource Loaded Schedule

    SciTech Connect (OSTI)

    BOWEN, W.W.

    1999-11-08

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FFTF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This document reflects the 1 Oct 1999 baseline.

  13. Advanced Reactors Transition Program Resource Loaded Schedule

    SciTech Connect (OSTI)

    GANTT, D.A.

    2000-01-12

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FETF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This revision reflects the 19 Oct 1999 baseline.

  14. Adaptive Nodal Transport Methods for Reactor Transient Analysis

    SciTech Connect (OSTI)

    Thomas Downar; E. Lewis

    2005-08-31

    Develop methods for adaptively treating the angular, spatial, and time dependence of the neutron flux in reactor transient analysis. These methods were demonstrated in the DOE transport nodal code VARIANT and the US NRC spatial kinetics code, PARCS.

  15. Heterogeneous Recycling in Fast Reactors

    SciTech Connect (OSTI)

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

    2012-07-30

    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.

  16. Advanced Safeguards Approaches for New Fast Reactors

    SciTech Connect (OSTI)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  17. On fast reactor kinetics studies

    SciTech Connect (OSTI)

    Seleznev, E. F.; Belov, A. A.; Matveenko, I. P.; Zhukov, A. M.; Raskach, K. F.

    2012-07-01

    The results and the program of fast reactor core time and space kinetics experiments performed and planned to be performed at the IPPE critical facility is presented. The TIMER code was taken as computation support of the experimental work, which allows transient equations to be solved in 3-D geometry with multi-group diffusion approximation. The number of delayed neutron groups varies from 6 to 8. The code implements the solution of both transient neutron transfer problems: a direct one, where neutron flux density and its derivatives, such as reactor power, etc, are determined at each time step, and an inverse one for the point kinetics equation form, where such a parameter as reactivity is determined with a well-known reactor power time variation function. (authors)

  18. LCLS Spectral Flux Viewer

    Energy Science and Technology Software Center (OSTI)

    2005-10-25

    This application (FluxViewer) is a tool for displaying spectral flux data for the Linac Coherent Light Source (LCLS). This tool allows the user to view sliced spatial and energy distributions of the photons selected for specific energies and positions transverse to the beam axis.

  19. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

    1958-09-01

    This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

  20. HIGS Flux Performance Projection

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

    HIGS flux performance table for high-flux, quasi-CW operation, DFELL/TUNL, Nov. 9, 2010 (Version 2.3). HIGS Flux Performance Projection (2010 - 2011) Total Flux [g/s] CW Operation Two-Bunch (*) Collimated Flux (∆E γ /E γ = 5% FWHM) (#), (@) FEL λ [nm] Comment No-loss Mode : < 20 MeV Linear Pol. with OK-4 Circular Pol with OK-5 E γ = 1 - 2 MeV (E e = 237 - 336 MeV) 1 x 10 8 - 4 x 10 8 6 x 10 6 - 2.4 x 10 7 1064 Linear and Circular (a), (b) E γ = 2 - 2.9 MeV (E e = 336 - 405 MeV) 4 x 10

  1. Generation of Radixenon Isotopes

    SciTech Connect (OSTI)

    McIntyre, Justin I.; Bowyer, Ted W.; Hayes, James C.; Heimbigner, Tom R.; Morris, Scott J.; Panisko, Mark E.; Pitts, W. K.; Pratt, Sharon L.; Reeder, Paul L.; Thomas, Charles W.

    2003-06-30

    Pacific Northwest National Laboratory has developed an automated system for separating Xe from air and can detect the following radioxenon isotopes, 131mXe, 133mXe, 133Xe, and 135Xe. This report details the techniques used to generate the various radioxenon isotopes that are used for the calibration of the detector as well as other isotopes that have the potential to interfere with the fission produced radioxenon isotopes. Fission production is covered first using highly enriched uranium followed by a description and results from an experiment to produce radioxenon isotopes from neutron activation of ambient xenon.

  2. Catalytic reactor

    DOE Patents [OSTI]

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    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.

  3. REACTOR CONTROL

    DOE Patents [OSTI]

    Fortescue, P.; Nicoll, D.

    1962-04-24

    A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

  4. Bioconversion reactor

    DOE Patents [OSTI]

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    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.

  5. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  6. Reactor physics methods, models, and applications used to support the conceptual design of the Advanced Neutron Source

    SciTech Connect (OSTI)

    Gehin, J.C.; Worley, B.A.; Renier, J.P.; Wemple, C.A.; Jahshan, S.N.; Ryskammp, J.M.

    1995-08-01

    This report summarizes the neutronics analysis performed during 1991 and 1992 in support of characterization of the conceptual design of the Advanced Neutron Source (ANS). The methods used in the analysis, parametric studies, and key results supporting the design and safety evaluations of the conceptual design are presented. The analysis approach used during the conceptual design phase followed the same approach used in early ANS evaluations: (1) a strong reliance on Monte Carlo theory for beginning-of-cycle reactor performance calculations and (2) a reliance on few-group diffusion theory for reactor fuel cycle analysis and for evaluation of reactor performance at specific time steps over the fuel cycle. The Monte Carlo analysis was carried out using the MCNP continuous-energy code, and the few- group diffusion theory calculations were performed using the VENTURE and PDQ code systems. The MCNP code was used primarily for its capability to model the reflector components in realistic geometries as well as the inherent circumvention of cross-section processing requirements and use of energy-collapsed cross sections. The MCNP code was used for evaluations of reflector component reactivity effects and of heat loads in these components. The code was also used as a benchmark comparison against the diffusion-theory estimates of key reactor parameters such as region fluxes, control rod worths, reactivity coefficients, and material worths. The VENTURE and PDQ codes were used to provide independent evaluations of burnup effects, power distributions, and small perturbation worths. The performance and safety calculations performed over the subject time period are summarized, and key results are provided. The key results include flux and power distributions over the fuel cycle, silicon production rates, fuel burnup rates, component reactivities, control rod worths, component heat loads, shutdown reactivity margins, reactivity coefficients, and isotope production rates.

  7. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Szilard, L.

    1957-09-24

    Reactors of the type employing plates of natural uranium in a moderator are discussed wherein the plates are um-formly disposed in parallel relationship to each other thereby separating the moderator material into distinct and individual layers. Each plate has an uninterrupted sunface area substantially equal to the cross-sectional area of the active portion of the reactor, the particular size of the plates and the volume ratio of moderator to uranium required to sustain a chain reaction being determinable from the known purity of these materials and other characteristics such as the predictable neutron losses due to the formation of radioactive elements of extremely high neutron capture cross section.

  8. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  9. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Young, G.

    1963-01-01

    This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

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

    SciTech Connect (OSTI)

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

    2010-04-01

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

  11. Atomic vapor laser isotope separation

    SciTech Connect (OSTI)

    Stern, R.C.; Paisner, J.A.

    1985-11-08

    Atomic vapor laser isotope separation (AVLIS) is a general and powerful technique. A major present application to the enrichment of uranium for light-water power reactor fuel has been under development for over 10 years. In June 1985 the Department of Energy announced the selection of AVLIS as the technology to meet the nation's future need for the internationally competitive production of uranium separative work. The economic basis for this decision is considered, with an indicated of the constraints placed on the process figures of merit and the process laser system. We then trace an atom through a generic AVLIS separator and give examples of the physical steps encountered, the models used to describe the process physics, the fundamental parameters involved, and the role of diagnostic laser measurements.

  12. Homopolar Gun for Pulsed Spheromak Fusion Reactors II

    SciTech Connect (OSTI)

    Fowler, T

    2004-06-14

    A homopolar gun is discussed that could produce the high currents required for pulsed spheromak fusion reactors even with unit current amplification and open field lines during injection, possible because close coupling between the gun and flux conserver reduces gun losses to acceptable levels. Example parameters are given for a gun compatible with low cost pulsed reactors and for experiments to develop the concept.

  13. Self-actuating reactor shutdown system

    DOE Patents [OSTI]

    Barrus, Donald M.; Brummond, Willian A; Peterson, Leslie F.

    1988-01-01

    A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

  14. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Starr, C.

    1963-01-01

    This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

  15. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

    1959-10-27

    BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

  16. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J.

    1976-08-17

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

  17. Neutronic reactor

    DOE Patents [OSTI]

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  18. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Anderson, H.L.

    1958-10-01

    The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

  19. ARM - Measurement - Actinic flux

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

    would love to hear from you Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Actinic flux The quantity of light in the atmosphere available to molecules at a...

  20. JACKETED FUEL ELEMENTS FOR GRAPHITE MODERATED REACTORS

    DOE Patents [OSTI]

    Szilard, L.; Wigner, E.P.; Creutz, E.C.

    1959-05-12

    Fuel elements for a heterogeneous, fluid cooled, graphite moderated reactor are described. The fuel elements are comprised of a body of natural uranium hermetically sealed in a jacket of corrosion resistant material. The jacket, which may be aluminum or some other material which is non-fissionable and of a type having a low neutron capture cross-section, acts as a barrier between the fissioning isotope and the coolant or moderator or both. The jacket minimizes the tendency of the moderator and coolant to become radioactive and/or contaminated by fission fragments from the fissioning isotope.

  1. HYDROGEN ISOTOPE TARGETS

    DOE Patents [OSTI]

    Ashley, R.W.

    1958-08-12

    The design of targets for use in the investigation of nuclear reactions of hydrogen isotopes by bombardment with accelerated particles is described. The target con struction eomprises a backing disc of a metal selected from the group consisting of molybdenunn and tungsten, a eoating of condensed titaniunn on the dise, and a hydrogen isotope selected from the group consisting of deuterium and tritium absorbed in the coatiag. The proeess for preparing these hydrogen isotope targets is described.

  2. Uranium Isotopic Assay Instrument

    SciTech Connect (OSTI)

    Anheier, Norman C.; Wojcik, Michael D.; Bushaw, Bruce A.

    2006-12-01

    The isotopic assay instrument under development at Pacific Northwest National Laboratory (PNNL) is capable of rapid prescreening to detect small and rare particles containing high concentrations of uranium in a heterogeneous sample. The isotopic measurement concept is based on laser vaporization of solid samples followed with sensitive isotope specific detection using either uranium atomic fluorescence emission or uranium atomic absorbance. Both isotopes are measured concurrently, following a single ablation laser pulse, using two external-cavity violet diode lasers. The simultaneous measurement of both isotopes enables the correlation of the fluorescence and absorbance signals on a shot-to-shot basis. This measurement approach demonstrated negligible channel crosstalk between isotopes. Rapid sample scanning provides high spatial resolution isotopic fluorescence and absorbance sample imagery of heterogeneous samples. Laser ablation combined with measurements of laser-induced fluorescence (LALIF) and through-plume laser absorbance (LAPLA) was applied to measure gadolinium isotope ratios in solid samples. Gadolinium has excitation wavelengths very close to the transitions of interest in uranium. Gadolinium has seven stable isotopes, and the natural 152Gd:160Gd ratio of 0.009 is in the range of what will be encountered for 235U:238U isotopic ratios. LAPLA measurements were demonstrated clearly using 152Gd (0.2% isotopic abundance) with a good signal-to-noise ratio. The ability to measure gadolinium abundances at this level indicates that measurements of 235U/238U isotopic ratios for natural (0.72%), depleted (0.25%), and low enriched uranium samples will be feasible.

  3. ARM - Measurement - Isotope ratio

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

    govMeasurementsIsotope ratio ARM Data Discovery Browse Data Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Isotope ratio Ratio of stable isotope concentrations. Categories Atmospheric State, Atmospheric Carbon Instruments The above measurement is considered scientifically relevant for the following instruments. Refer to the datastream (netcdf) file headers of each instrument for a list of all available measurements, including those

  4. Hybrid isotope separation scheme

    DOE Patents [OSTI]

    Maya, J.

    1991-06-18

    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.

  5. Hybrid isotope separation scheme

    DOE Patents [OSTI]

    Maya, Jakob

    1991-01-01

    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.

  6. Stable isotope enrichment

    ScienceCinema (OSTI)

    Egle, Brian

    2014-07-15

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

  7. Stable isotope enrichment

    SciTech Connect (OSTI)

    Egle, Brian

    2014-07-14

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

  8. REACTOR MONITORING

    DOE Patents [OSTI]

    Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

    1959-02-01

    A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

  9. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Treshow, M.

    1958-08-19

    A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

  10. REACTOR UNLOADING

    DOE Patents [OSTI]

    Leverett, M.C.

    1958-02-18

    This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

  11. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wade, E.J.

    1958-09-16

    This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

  12. Nuclear reactor

    DOE Patents [OSTI]

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  13. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  14. NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  15. Neutronic reactor

    DOE Patents [OSTI]

    Lewis, Warren R.

    1978-05-30

    A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

  16. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  17. REACTOR CONTROL

    DOE Patents [OSTI]

    Ruano, W.J.

    1957-12-10

    This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

  18. Isotopic Generation and Confirmation of the PWR Application Model 

    SciTech Connect (OSTI)

    L.B. Wimmer

    2003-11-10

    The objective of this calculation is to establish an isotopic database to represent commercial spent nuclear fuel (CSNF) from pressurized water reactors (PWRs) in criticality analyses performed for the proposed Monitored Geologic Repository at Yucca Mountain, Nevada. Confirmation of the conservatism with respect to criticality in the isotopic concentration values represented by this isotopic database is performed as described in Section 3.5.3.1.2 of the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000). The isotopic database consists of the set of 14 actinides and 15 fission products presented in Section 3.5.2.1.1 of YMP 2000 for use in CSNF burnup credit. This set of 29 isotopes is referred to as the principal isotopes. The oxygen isotope from the UO{sub 2} fuel is also included in the database. The isotopic database covers enrichments of {sup 235}U ranging from 1.5 to 5.5 weight percent (wt%) and burnups ranging from approximately zero to 75 GWd per metric ton of uranium (mtU). The choice of fuel assembly and operating history values used in generating the isotopic database are provided is Section 5. Tables of isotopic concentrations for the 29 principal isotopes (plus oxygen) as a function of enrichment and burnup are provided in Section 6.1. Results of the confirmation of the conservatism with respect to criticality in the isotopic concentration values are provided in Section 6.2.

  19. Cumulative fission yields of short-lived isotopes under natural-abundance-boron-carbide-moderated neutron spectrum

    SciTech Connect (OSTI)

    Finn, Erin C.; Metz, Lori A.; Greenwood, Lawrence R.; Pierson, Bruce; Wittman, Richard S.; Friese, Judah I.; Kephart, Rosara F.

    2015-04-09

    The availability of gamma spectroscopy data on samples containing mixed fission products at short times after irradiation is limited. Due to this limitation, data interpretation methods for gamma spectra of mixed fission product samples, where the individual fission products have not been chemically isolated from interferences, are not well-developed. The limitation is particularly pronounced for fast pooled neutron spectra because of the lack of available fast reactors in the United States. Samples containing the actinide isotopes 233, 235, 238U, 237Np, and 239Pu individually were subjected to a 2$ pulse in the Washington State University 1 MW TRIGA reactor. To achieve a fission-energy neutron spectrum, the spectrum was tailored using a natural abundance boron carbide capsule to absorb neutrons in the thermal and epithermal region of the spectrum. Our tailored neutron spectrum is unique to the WSU reactor facility, consisting of a soft fission spectrum that contains some measurable flux in the resonance region. This results in a neutron spectrum at greater than 0.1 keV with an average energy of 70 keV, similar to fast reactor spectra and approaching that of 235U fission. Unique fission product gamma spectra were collected from 4 minutes to 1 week after fission using single-crystal high purity germanium detectors. Cumulative fission product yields measured in the current work generally agree with published fast pooled fission product yield values from ENDF/B-VII, though a bias was noted for 239Pu. The present work contributes to the compilation of energy-resolved fission product yield nuclear data for nuclear forensic purposes.

  20. Price Quotes and Isotope Ordering

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

    Ordering Price Quotes and Isotope Ordering Isotopes produced at Los Alamos National Laboratory are saving lives, advancing cutting-edge research and keeping the U.S. safe. Isotope...

  1. Uranium accountancy in Atomic Vapor Laser Isotope Separation

    SciTech Connect (OSTI)

    Carver, R.D.

    1986-01-01

    The AVLIS program pioneers the large scale industrial application of lasers to produce low cost enriched uranium fuel for light water reactors. In the process developed at Lawrence Livermore National Laboratory, normal uranium is vaporized by an electron beam, and a precisely tuned laser beam selectively photo-ionizes the uranium-235 isotopes. These ions are moved in an electromagnetic field to be condensed on the product collector. All other uranium isotopes remain uncharged and pass through the collector section to condense as tails. Tracking the three types of uranium through the process presents special problems in accountancy. After demonstration runs, the uranium on the collector was analyzed for isotopic content by Battelle Pacific Northwest Laboratory. Their results were checked at LLNL by analysis of parallel samples. The differences in isotopic composition as reported by the two laboratories were not significant.

  2. Advanced Test Reactor - A National Scientific User Facility

    SciTech Connect (OSTI)

    Clifford J. Stanley

    2008-05-01

    The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected nuclear research reactor with a maximum operating power of 250 MWth. The unique serpentine configuration of the fuel elements creates five main reactor power lobes (regions) and nine flux traps. In addition to these nine flux traps there are 68 additional irradiation positions in the reactor core reflector tank. There are also 34 low-flux irradiation positions in the irradiation tanks outside the core reflector tank. The ATR is designed to provide a test environment for the evaluation of the effects of intense radiation (neutron and gamma). Due to the unique serpentine core design each of the five lobes can be operated at different powers and controlled independently. Options exist for the individual test trains and assemblies to be either cooled by the ATR coolant (i.e., exposed to ATR coolant flow rates, pressures, temperatures, and neutron flux) or to be installed in their own independent test loops where such parameters as temperature, pressure, flow rate, neutron flux, and energy can be controlled per experimenter specifications. The full-power maximum thermal neutron flux is ~1.0 x1015 n/cm2-sec with a maximum fast flux of ~5.0 x1014 n/cm2-sec. The Advanced Test Reactor, now a National Scientific User Facility, is a versatile tool in which a variety of nuclear reactor, nuclear physics, reactor fuel, and structural material irradiation experiments can be conducted. The cumulative effects of years of irradiation in a normal power reactor can be duplicated in a few weeks or months in the ATR due to its unique design, power density, and operating flexibility.

  3. Quantum flux parametron

    SciTech Connect (OSTI)

    Hioe, W. ); Goto, E. )

    1991-01-01

    The quantum flux parametron (QFP) is an offspring of the parametron, an early flux-based logic device, and the Josephson junction. It is a single flux quantum device that works completely in the superconductive mode. While it has the speed of other Josephson devices that work on switching between the voltage and superconductive modes, its power is about one thousand times less. Hence, it promises to be an attractive alternative to both transistors and other Josephson devices. This book reports the latest research results on QFP applications as a logic device. In particular, a number of auxiliary circuits and a new logic gate are proposed for improving the device margin. Samples of these circuits and logic gate have been fabricated.

  4. 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

    2012-01-01

    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).

  5. Photochemical isotope separation

    DOE Patents [OSTI]

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

    1987-04-28

    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.

  6. Laser isotope separation

    DOE Patents [OSTI]

    Robinson, C. Paul; Jensen, Reed J.; Cotter, Theodore P.; Boyer, Keith; Greiner, Norman R.

    1988-01-01

    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.

  7. Photochemical isotope separation

    DOE Patents [OSTI]

    Robinson, C. Paul; Jensen, Reed J.; Cotter, Theodore P.; Greiner, Norman R.; Boyer, Keith

    1987-01-01

    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.

  8. Laser isotope separation

    DOE Patents [OSTI]

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

    1975-11-26

    A process and apparatus for separating isotopes by selective excitation of isotopic species of a volatile compound by tuned laser light is described. 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.

  9. B Reactor | Department of Energy

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

    Operational Management History Manhattan Project Signature Facilities B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first ...

  10. Nuclear reactor

    DOE Patents [OSTI]

    Pennell, William E.; Rowan, William J.

    1977-01-01

    A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

  11. NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Ashley, J.W.

    1958-12-16

    A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

  12. Photocatalytic reactor

    DOE Patents [OSTI]

    Bischoff, B.L.; Fain, D.E.; Stockdale, J.A.D.

    1999-01-19

    A photocatalytic reactor is described for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane. 4 figs.

  13. Radiative Flux Analysis

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

    Long, Chuck [NOAA

    2008-05-14

    The Radiative Flux Analysis is a technique for using surface broadband radiation measurements for detecting periods of clear (i.e. cloudless) skies, and using the detected clear-sky data to fit functions which are then used to produce continuous clear-sky estimates. The clear-sky estimates and measurements are then used in various ways to infer cloud macrophysical properties.

  14. Slide 1

    Office of Environmental Management (EM)

    ... environmental sciences, biosciences, medical sciences, and pharmaceutical sciences. * ... Neutron Source, ORNL * High Flux Isotope Reactor, ORNL (Top) Spallation neutron ...

  15. Hybrid adsorptive membrane reactor

    DOE Patents [OSTI]

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    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.

  16. H Reactor - Hanford Site

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

    About Us Projects & Facilities H Reactor About Us About Hanford Cleanup Hanford History ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  17. C Reactor - Hanford Site

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

    C Reactor About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  18. F Reactor - Hanford Site

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

    About Us Projects & Facilities F Reactor About Us About Hanford Cleanup Hanford History ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  19. N Reactor - Hanford Site

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

    Projects & Facilities N Reactor About Us About Hanford Cleanup Hanford History Hanford ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  20. Power flow control using distributed saturable reactors

    DOE Patents [OSTI]

    Dimitrovski, Aleksandar D.

    2016-02-13

    A magnetic amplifier includes a saturable core having a plurality of legs. Control windings wound around separate legs are spaced apart from each other and connected in series in an anti-symmetric relation. The control windings are configured in such a way that a biasing magnetic flux arising from a control current flowing through one of the plurality of control windings is substantially equal to the biasing magnetic flux flowing into a second of the plurality of control windings. The flow of the control current through each of the plurality of control windings changes the reactance of the saturable core reactor by driving those portions of the saturable core that convey the biasing magnetic flux in the saturable core into saturation. The phasing of the control winding limits a voltage induced in the plurality of control windings caused by a magnetic flux passing around a portion of the saturable core.

  1. Science on Tap - Isotopes

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

    Science on Tap - Isotopes Science on Tap - Isotopes WHEN: Jun 16, 2016 5:30 PM - 7:00 PM WHERE: UnQuarked Wine Room 145 Central Park Square, Los Alamos, New Mexico 87544 USA CONTACT: Linda Anderman (505) 665-9196 CATEGORY: Bradbury INTERNAL: Calendar Login Event Description Short presentation followed by lively interaction on the topic at hand. While isotopes are chemical elements (think periodic table), their varying numbers of neutrons mean they can be used in lots of different way. Join us

  2. Power Burst Facility (PBF) Reactor Reactor Decommissioning

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

    Click here to view Click here to view Reactor Decommissioning Click on an image to enlarge A crane removes the reactor vessel from the Power Burst Facility (top), then places it ...

  3. Atomic vapor laser isotope separation

    SciTech Connect (OSTI)

    Stern, R.C.; Paisner, J.A.

    1986-08-15

    The atomic vapor laser isotope separation (AVLIS) process for the enrichment of uranium is evaluated. (AIP)

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

    DOE Patents [OSTI]

    Grossman, M.W.

    1991-04-30

    The present invention is directed to an improved photochemical reactor useful for the isotopic enrichment of a predetermined isotope of mercury, especially, [sup 196]Hg. Specifically, two axi-symmetrical flow reactors were constructed according to the teachings of the present invention. These reactors improve the mixing of the reactants during the photochemical enrichment process, affording higher yields of the desired [sup 196]Hg product. Measurements of the variation of yield (Y) and enrichment factor (E) along the flow axis of these reactors indicates very substantial improvement in process uniformity compared to previously used photochemical reactor systems. In one preferred embodiment of the present invention, the photoreactor system was built such that the reactor chamber was removable from the system without disturbing the location of either the photochemical lamp or the filter employed therewith. 10 figures.

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

    DOE Patents [OSTI]

    Grossman, Mark W.

    1991-01-01

    The present invention is directed to an improved photochemical reactor useful for the isotopic enrichment of a predetermined isotope of mercury, especially, .sup.196 Hg. Specifically, two axi-symmetrical flow reactors were constructed according to the teachings of the present invention. These reactors improve the mixing of the reactants during the photochemical enrichment process, affording higher yields of the desired .sup.196 Hg product. Measurements of the variation of yield (Y) and enrichment factor (E) along the flow axis of these reactors indicates very substantial improvement in process uniformity compared to previously used photochemical reactor systems. In one preferred embodiment of the present invention, the photoreactor system was built such that the reactor chamber was removable from the system without disturbing the location of either the photochemical lamp or the filter employed therewith.

  6. Atomic vapor laser isotope separation using resonance ionization

    SciTech Connect (OSTI)

    Comaskey, B.; Crane, J.; Erbert, G.; Haynam, C.; Johnson, M.; Morris, J.; Paisner, J.; Solarz, R.; Worden, E.

    1986-09-01

    Atomic vapor laser isotope separation (AVLIS) is a general and powerful technique. A major present application to the enrichment of uranium for light-water power-reactor fuel has been under development for over 10 years. In June 1985, the Department of Energy announced the selection of AVLIS as the technology to meet the nation's future need for enriched uranium. Resonance photoionization is the heart of the AVLIS process. We discuss those fundamental atomic parameters that are necessary for describing isotope-selective resonant multistep photoionization along with the measurement techniques that we use. We illustrate the methodology adopted with examples of other elements that are under study in our program.

  7. Control system for a small fission reactor

    DOE Patents [OSTI]

    Burelbach, James P.; Kann, William J.; Saiveau, James G.

    1986-01-01

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired. In another embodiment, a plurality of flexible hollow tubes each containing a neutron absorber are positioned adjacent to one another in spaced relation around the periphery of the reactor vessel and inside the outer neutron reflector with reactivity controlled by the extension and compression of all or some of the coiled hollow tubes. Yet another embodiment of the invention envisions the neutron reflector in the form of an expandable coil spring positioned in an annular space between the reactor vessel and an outer neutron absorbing structure for controlling the neutron flux reflected back into the reactor vessel.

  8. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  9. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Daniels, F.

    1962-12-18

    A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

  10. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wigner, E.P.

    1957-09-17

    A reactor of the type having coolant liquid circulated through clad fuel elements geometrically arranged in a solid moderator, such as graphite, is described. The core is enclosed in a pressure vessel and suitable shielding, wherein means is provided for circulating vapor through the core to superheat the same. This is accomplished by drawing off the liquid which has been heated in the core due to the fission of the fuel, passing it to a nozzle within a chamber where it flashes into a vapor, and then passing the vapor through separate tubes extending through the moderator to pick up more heat developed in the core due to the fission of the fuel, thereby producing superheated vapor.

  11. Nuclear reactor

    DOE Patents [OSTI]

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  12. SAFEGUARD AND SECURE CONTROL VERIFY POLICY

    National Nuclear Security Administration (NNSA)

    ... Ridge National Laboratory where nuclear isotopes are manipulated and processed for medical ... as tours and lectures at the High Flux Isotope Reactor (HFIR) and ORNL's Radiochemical ...

  13. Optical heat flux gauge

    DOE Patents [OSTI]

    Noel, B.W.; Borella, H.M.; Cates, M.R.; Turley, W.D.; MacArthur, C.D.; Cala, G.C.

    1991-04-09

    A heat flux gauge is disclosed comprising first and second thermographic phosphor layers separated by a layer of a thermal insulator, wherein each thermographic layer comprises a plurality of respective thermographic sensors in a juxtaposed relationship with respect to each other. The gauge may be mounted on a surface with the first thermographic phosphor in contact with the surface. A light source is directed at the gauge, causing the phosphors to luminesce. The luminescence produced by the phosphors is collected and its spectra analyzed in order to determine the heat flux on the surface. First and second phosphor layers must be different materials to assure that the spectral lines collected will be distinguishable. 9 figures.

  14. IN-CORE FLUX SENSOR EVALUATIONS AT THE ATR CRITICAL FACILITY.

    SciTech Connect (OSTI)

    Troy Unruh; Benjamin Chase; Joy Rempe; David Nigg; George Imel; Jason Harris; Todd Sherman; Jean-Francois VIllard

    2014-12-01

    As part of an Idaho State University (ISU)–led Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) collaborative project that includes Idaho National Laboratory (INL) and the French Alternative Energies and Atomic Energy Commission (CEA), flux detector evaluations were completed to compare their accuracy, response time, and longduration performance. Special fixturing, developed by INL, allows real-time flux detectors to be inserted into various Advanced Test Reactor Critical Facility (ATRC) core positions to perform lobe power measurements, axial flux profile measurements, and detector crosscalibrations. Detectors initially evaluated in this program included miniature fission chambers, specialized self-powered neutron detectors (SPNDs), and specially developed commercial SPNDs. Results from this program provide important insights related to flux detector accuracy and resolution for subsequent ATR and CEA experiments and yield new flux data required for benchmarking models in the ATR Life Extension Program (LEP) Modeling Update Project.

  15. Plasma isotope separation methods

    SciTech Connect (OSTI)

    Grossman, M.W. ); Shepp, T.A. )

    1991-12-01

    Isotope separation has many important industrial, medical, and research applications. Large-scale processes have typically utilized complex cascade systems; for example, the gas centrifuge. Alternatively, high single-stage enrichment processes (as in the case of the calutron) are very energy intensive. Plasma-based methods being developed for the past 15 to 20 years have attempted to overcome these two drawbacks. In this review, six major types of isotope separation methods which involve plasma phenomena are discussed. These methods are: plasma centrifuge, AVLIS (atomic vapor laser isotope separation), ion wave, ICR (ion-cyclotron resonance), calutron, and gas discharge. The emphasis of this paper is to describe the plasma phenomena in these major categories. An attempt was made to include enough references so that more detailed study or evaluation of a particular method could readily be pursued. A brief discussion of isotope separation using mass balance concepts is also carried out.

  16. Tritium Formation and Mitigation in High Temperature Reactors

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots

    2012-08-01

    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.

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

    SciTech Connect (OSTI)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    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

  18. Health and safety consequences of medical isotope processing at the Hanford Site 325 building

    SciTech Connect (OSTI)

    Nielsen, D. L.

    1997-11-19

    Potential activities associated with medical isotope processing at the Hanford Site 325 Building laboratory and hot cell facilities are evaluated to assess the health and safety consequences if these activities are to be implemented as part of a combined tritium and medical isotope production mission for the Fast Flux Text Facility (FFTF). The types of activities included in this analysis are unloading irradiated isotope production assemblies at the 325 Building, recovery and dissolution of the target materials, separation of the product isotopes as required, and preparation of the isotopes for shipment to commercial distributors who supply isotopes to the medical conunuriity. Possible consequences to members of the public and to workers from both radiological and non-radiological hazards are considered in this evaluation. Section 2 of this docinnent describes the assumptions and methods used for the health and safety consequences analysis, section 3 presents the results of the analysis, and section 4 summarizes the results and conclusions from the analysis.

  19. Reactor and method of operation

    DOE Patents [OSTI]

    Wheeler, John A.

    1976-08-10

    A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

  20. Effect of rolling motion on critical heat flux for subcooled flow boiling in vertical tube

    SciTech Connect (OSTI)

    Hwang, J. S.; Park, I. U.; Park, M. Y.; Park, G. C.

    2012-07-01

    This paper presents defining characteristics of the critical heat flux (CHF) for the boiling of R-134a in vertical tube operation under rolling motion in marine reactor. It is important to predict CHF of marine reactor having the rolling motion in order to increase the safety of the reactor. Marine Reactor Moving Simulator (MARMS) tests are conducted to measure the critical heat flux using R-134a flowing upward in a uniformly heated vertical tube under rolling motion. MARMS was rotated by motor and mechanical power transmission gear. The CHF tests were performed in a 9.5 mm I.D. test section with heated length of 1 m. Mass fluxes range from 285 to 1300 kg m{sup -2}s{sup -1}, inlet subcooling from 3 to 38 deg. C and outlet pressures from 13 to 24 bar. Amplitudes of rolling range from 15 to 40 degrees and periods from 6 to 12 sec. To convert the test conditions of CHF test using R-134a in water, Katto's fluid-to-fluid modeling was used in present investigation. A CHF correlation is presented which accounts for the effects of pressure, mass flux, inlet subcooling and rolling angle over all conditions tested. Unlike existing transient CHF experiments, CHF ratio of certain mass flux and pressure are different in rolling motion. For the mass fluxes below 500 kg m{sup -2}s{sup -1} at 13, 16 (region of relative low mass flux), CHF ratio was decreased but was increased above that mass flux (region of relative high mass flux). Moreover, CHF tend to enhance in entire mass flux at 24 bar. (authors)

  1. ARM - Measurement - Sensible heat flux

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

    heat flux ARM Data Discovery Browse Data Comments? We would love to hear from you Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Sensible heat flux The time ...

  2. ARM - Measurement - Latent heat flux

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

    heat flux ARM Data Discovery Browse Data Comments? We would love to hear from you Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Latent heat flux The time ...

  3. Light Water Reactor Sustainability (LWRS) Program | Department...

    Energy Savers [EERE]

    Nuclear Reactor Technologies Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) ...

  4. SRS Small Modular Reactors

    ScienceCinema (OSTI)

    None

    2014-05-21

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

  5. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J.

    1980-03-11

    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.

  6. Hydrogen isotope separation utilizing bulk getters

    DOE Patents [OSTI]

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

    1990-01-01

    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.

  7. Hydrogen isotope separation utilizing bulk getters

    DOE Patents [OSTI]

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

    1991-01-01

    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.

  8. Hydrogen isotope separation utilizing bulk getters

    DOE Patents [OSTI]

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

    1991-08-20

    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.

  9. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B.

    2004-03-16

    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.

  10. Fast Flux Test Facility Closure Project - Project Management Plan

    SciTech Connect (OSTI)

    BEACH, R.R.

    2002-09-26

    The Fast Flux Test Facility (FFTF) Closure Project, Project Management Plan, Revision 5, provides the scope, cost, and schedule to achieve the most cost effective and expeditious closure of the FFTF to an assumed final end-state with the reactor vessel and the containment building, below the 5504 grade level, being entombed in place. Closure will be completed by December 2009 at a cost of $547 million.

  11. In the OSTI Collections: Neutron Sources for Studying Matter...

    Office of Scientific and Technical Information (OSTI)

    ... by removing preexisting neutrons from nuclei using any of several nuclear reactions. ... The High Flux Isotope Reactor works like any nuclear reactor that involves a chain of ...

  12. PIK M.S. Onegin Petersburg Nuclear Physics Institute 2015 Super...

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

    isotopes production in high-flux reactor PIK M.S. Onegin Petersburg Nuclear Physics Institute 2015 Super Heavy Elements Symposium Reactor PIK 2011 - Criticality reached 2013 - ...

  13. Prospects for Tokamak Fusion Reactors

    SciTech Connect (OSTI)

    Sheffield, J.; Galambos, J.

    1995-04-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.

  14. Separation of sulfur isotopes

    DOE Patents [OSTI]

    DeWitt, Robert; Jepson, Bernhart E.; Schwind, Roger A.

    1976-06-22

    Sulfur isotopes are continuously separated and enriched using a closed loop reflux system wherein sulfur dioxide (SO.sub.2) is reacted with sodium hydroxide (NaOH) or the like to form sodium hydrogen sulfite (NaHSO.sub.3). Heavier sulfur isotopes are preferentially attracted to the NaHSO.sub.3, and subsequently reacted with sulfuric acid (H.sub.2 SO.sub.4) forming sodium hydrogen sulfate (NaHSO.sub.4) and SO.sub.2 gas which contains increased concentrations of the heavier sulfur isotopes. This heavy isotope enriched SO.sub.2 gas is subsequently separated and the NaHSO.sub.4 is reacted with NaOH to form sodium sulfate (Na.sub.2 SO.sub.4) which is subsequently decomposed in an electrodialysis unit to form the NaOH and H.sub.2 SO.sub.4 components which are used in the aforesaid reactions thereby effecting sulfur isotope separation and enrichment without objectionable loss of feed materials.

  15. Isotope separation apparatus

    DOE Patents [OSTI]

    Arnush, Donald; MacKenzie, Kenneth R.; Wuerker, Ralph F.

    1980-01-01

    Isotope separation apparatus consisting of a plurality of cells disposed adjacent to each other in an evacuated container. A common magnetic field is established extending through all of the cells. A source of energetic electrons at one end of the container generates electrons which pass through the cells along the magnetic field lines. Each cell includes an array of collector plates arranged in parallel or in tandem within a common magnetic field. Sets of collector plates are disposed adjacent to each other in each cell. Means are provided for differentially energizing ions of a desired isotope by applying energy at the cyclotron resonant frequency of the desired isotope. As a result, the energized desired ions are preferentially collected by the collector plates.

  16. Breeding of {sup 233}U in the thorium–uranium fuel cycle in VVER reactors using heavy water

    SciTech Connect (OSTI)

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

    2015-12-15

    A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the {sup 233}U–{sup 232}Th oxide fuel of water-moderated reactors with variable water composition (D{sub 2}O, H{sub 2}O) that ensures breeding of the {sup 233}U and {sup 235}U isotopes. The method is comparatively simple to implement.

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

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

    Department of Energy Research and Development Roadmap: Windows and Building Envelope Research and Development Roadmap: Windows and Building Envelope Cover of windows and envelope report, depicting a house, storefront, and multiple office windows. This Building Technologies Office (BTO) Research and Development (R&D) Roadmap identifies priority windows and building envelope R&D areas of interest. Cost and performance targets are identified for each key R&D area. The roadmap

  18. SPRAY CALCINATION REACTOR

    DOE Patents [OSTI]

    Johnson, B.M.

    1963-08-20

    A spray calcination reactor for calcining reprocessin- g waste solutions is described. Coaxial within the outer shell of the reactor is a shorter inner shell having heated walls and with open regions above and below. When the solution is sprayed into the irner shell droplets are entrained by a current of gas that moves downwardly within the inner shell and upwardly between it and the outer shell, and while thus being circulated the droplets are calcined to solids, whlch drop to the bottom without being deposited on the walls. (AEC) H03 H0233412 The average molecular weights of four diallyl phthalate polymer samples extruded from the experimental rheometer were redetermined using the vapor phase osmometer. An amine curing agent is required for obtaining suitable silver- filled epoxy-bonded conductive adhesives. When the curing agent was modified with a 47% polyurethane resin, its effectiveness was hampered. Neither silver nor nickel filler impart a high electrical conductivity to Adiprenebased adhesives. Silver filler was found to perform well in Dow-Corning A-4000 adhesive. Two cascaded hot-wire columns are being used to remove heavy gaseous impurities from methane. This purified gas is being enriched in the concentric tube unit to approximately 20% carbon-13. Studies to count low-level krypton-85 in xenon are continuing. The parameters of the counting technique are being determined. The bismuth isotopes produced in bismuth irradiated for polonium production are being determined. Preliminary data indicate the presence of bismuth207 and bismuth-210m. The light bismuth isotopes are probably produced by (n,xn) reactions bismuth-209. The separation of uranium-234 from plutonium-238 solutions was demonstrated. The bulk of the plutonium is removed by anion exchange, and the remainder is extracted from the uranium by solvent extraction techniques. About 99% of the plutonium can be removed in each thenoyltrifluoroacetone extraction. The viscosity, liquid density, and

  19. Control Rod Malfunction at the NRAD Reactor

    SciTech Connect (OSTI)

    Thomas L. Maddock

    2010-05-01

    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.

  20. FAST FLUX TEST FACILITY (FFTF) A HISTORY OF SAFETY & OPERATIONAL EXCELLENCE

    SciTech Connect (OSTI)

    NIELSEN, D L

    2004-02-26

    The Fast Flux Test Facility (FFTF) is a 400-megawatt (thermal) sodium-cooled, high temperature, fast neutron flux, loop-type test reactor. The facility was constructed to support development and testing of fuels, materials and equipment for the Liquid Metal Fast Breeder Reactor program. FFTF began operation in 1980 and over the next 10 years demonstrated its versatility to perform experiments and missions far beyond the original intent of its designers. The reactor had several distinctive features including its size, flux, core design, extensive instrumentation, and test features that enabled it to simultaneously carry out a significant array of missions while demonstrating its features that contributed to a high level of plant safety and availability. FFTF is currently being deactivated for final closure.

  1. BDDR, a new CEA technological and operating reactor database

    SciTech Connect (OSTI)

    Soldevilla, M.; Salmons, S.; Espinosa, B.

    2013-07-01

    The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

  2. DEEP WATER ISOTOPIC CURRENT ANALYZER

    DOE Patents [OSTI]

    Johnston, W.H.

    1964-04-21

    A deepwater isotopic current analyzer, which employs radioactive isotopes for measurement of ocean currents at various levels beneath the sea, is described. The apparatus, which can determine the direction and velocity of liquid currents, comprises a shaft having a plurality of radiation detectors extending equidistant radially therefrom, means for releasing radioactive isotopes from the shaft, and means for determining the time required for the isotope to reach a particular detector. (AEC)

  3. Method for separating boron isotopes

    DOE Patents [OSTI]

    Rockwood, Stephen D.

    1978-01-01

    A method of separating boron isotopes .sup.10 B and .sup.11 B by laser-induced selective excitation and photodissociation of BCl.sub.3 molecules containing a particular boron isotope. The photodissociation products react with an appropriate chemical scavenger and the reaction products may readily be separated from undissociated BCl.sub.3, thus effecting the desired separation of the boron isotopes.

  4. LIGHT WATER MODERATED NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Christy, R.F.; Weinberg, A.M.

    1957-09-17

    A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

  5. Once-through CANDU reactor models for the ORIGEN2 computer code

    SciTech Connect (OSTI)

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % /sup 235/U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given.

  6. Attrition reactor system

    DOE Patents [OSTI]

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

    1993-09-28

    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.

  7. Attrition reactor system

    DOE Patents [OSTI]

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

    1993-01-01

    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.

  8. Hybrid plasmachemical reactor

    SciTech Connect (OSTI)

    Lelevkin, V. M. Smirnova, Yu. G.; Tokarev, A. V.

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  9. Further investigation of the “reactor anomaly”

    SciTech Connect (OSTI)

    Garvey, G. T. Hayes, A. C. Jungman, Gerard; Jonkmans, G.

    2015-07-15

    The effect of a more realistic and extensive inclusion of first forbidden beta decay into the determination of the reactor neutrino flux is investigated. Forbidden decays make up approximately 30% of all fission product decays so their possible impact on the neutrino flux should not be neglected. Because of an incomplete knowledge of the requisite nuclear structure it is not possible to incorporate the forbidden decays in an exact fashion thus a variety of scenarios are investigated. It appears that including first forbidden decays measurably modifies the anti-neutrino spectrum, and the uncertainty on the neutrino flux should be expanded beyond 4%.

  10. Reactor System Transient Code.

    Energy Science and Technology Software Center (OSTI)

    1999-07-14

    RELAP3B describes the behavior of water-cooled nuclear reactors during postulated accidents or power transients, such as large reactivity excursions, coolant losses or pump failures. The program calculates flows, mass and energy inventories, pressures, temperatures, and steam qualities along with variables associated with reactor power, reactor heat transfer, or control systems. Its versatility allows one to describe simple hydraulic systems as well as complex reactor systems.

  11. Period meter for reactors

    DOE Patents [OSTI]

    Rusch, Gordon K.

    1976-01-06

    An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

  12. NEUTRONIC REACTOR POWER PLANT

    DOE Patents [OSTI]

    Metcalf, H.E.

    1962-12-25

    This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

  13. Preserving physics knowledge at the fast flux test facility

    SciTech Connect (OSTI)

    Wootan, D.; Omberg, R.; Makenas, B. J.; Polzin, D. L.

    2012-07-01

    One of the goals of the Dept. of Energy's Office of Nuclear Energy, initiated under the Fuel Cycle Research and Development Program (FCRD) and continued under the Advanced Reactor Concepts Program (ARC) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMRs) that could support the development of an environmentally and economically sound nuclear fuel cycle. The Fast Flux Test Facility (FFTF) is the most recent LMR to operate in the United States, from 1982 to 1992, and was designed as a fully instrumented test reactor with on-line, real time test control and performance monitoring of components and tests installed in the reactor. The 10 years of operation of the FFTF provided a very useful framework for testing the advances in LMR safety technology based on passive safety features that may be of increased importance to new designs after the events at Fukushima. Knowledge preservation at the FFTF is focused on the areas of design, construction, and startup of the reactor, as well as on preserving information obtained from 10 years of successful operating history and extensive irradiation testing of fuels and materials. In order to ensure protection of information at risk, the program to date has sequestered reports, files, tapes, and drawings to allow for secure retrieval. A disciplined and orderly approach has been developed to respond to client's requests for documents and data in order to minimize the search effort and ensure that future requests for this information can be readily accommodated. (authors)

  14. NEUTRONIC REACTOR SHIELDING

    DOE Patents [OSTI]

    Borst, L.B.

    1961-07-11

    A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

  15. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28

    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.

  16. High solids fermentation reactor

    DOE Patents [OSTI]

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

    1993-01-01

    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.

  17. High solids fermentation reactor

    DOE Patents [OSTI]

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

    1993-03-02

    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. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, James P.; Scahill, John W.

    1995-01-01

    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.

  19. Advanced Test Reactor Tour

    SciTech Connect (OSTI)

    Miley, Don

    2011-01-01

    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.

  20. NUCLEAR REACTOR CONTROL SYSTEM

    DOE Patents [OSTI]

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

  1. CRDIAC: Coupled Reactor Depletion Instrument with Automated Control

    SciTech Connect (OSTI)

    Steven K. Logan

    2012-08-01

    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

  2. Laser-isotope-separation technology. [Review; economics

    SciTech Connect (OSTI)

    Jensen, R.J.; Blair, L.S.

    1981-01-01

    The Molecular Laser Isotope Separation (MLIS) process currently under development is discussed as an operative example of the use of lasers for material processing. The MLIS process, which uses infrared and ultraviolet lasers to process uranium hexafluoride (UF/sub 6/) resulting in enriched uranium fuel to be used in electrical-power-producing nuclear reactor, is reviewed. The economics of the MLIS enrichment process is compared with conventional enrichment technique, and the projected availability of MLIS enrichment capability is related to estimated demands for U.S. enrichment service. The lasers required in the Los Alamos MLIS program are discussed in detail, and their performance and operational characteristics are summarized. Finally, the timely development of low-cost, highly efficient ultraviolet and infrared lasers is shownd to be the critical element controlling the ultimate deployment of MLIS uranium enrichment. 8 figures, 7 tables.

  3. PHLUX: Photographic Flux Tools for Solar Glare and Flux

    Energy Science and Technology Software Center (OSTI)

    2010-12-02

    A web-based tool to a) analytically and empirically quantify glare from reflected light and determine the potential impact (e.g., temporary flash blindness, retinal burn), and b) produce flux maps for central receivers. The tool accepts RAW digital photographs of the glare source (for hazard assessment) or the receiver (for flux mapping), as well as a photograph of the sun for intensity and size scaling. For glare hazard assessment, the tool determines the retinal irradiance (W/cm2)more » and subtended source angle for an observer and plots the glare source on a hazard spectrum (i.e., low-potential for flash blindness impact, potential for flash blindness impact, retinal burn). For flux mapping, the tool provides a colored map of the receiver scaled by incident solar flux (W/m2) and unwraps the physical dimensions of the receiver while accounting for the perspective of the photographer (e.g., for a flux map of a cylindrical receiver, the horizontal axis denotes receiver angle in degrees and the vertical axis denotes vertical position in meters; for a flat panel receiver, the horizontal axis denotes horizontal position in meters and the vertical axis denotes vertical position in meters). The flux mapping capability also allows the user to specify transects along which the program plots incident solar flux on the receiver.« less

  4. Flux growth utilizing the reaction between flux and crucible

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

    Yan, J. -Q.

    2015-01-22

    Flux growth involves dissolving the components of the target compound in an appropriate flux at high temperatures and then crystallizing under supersaturation controlled by cooling or evaporating the flux. A refractory crucible is generally used to contain the high temperature melt. Moreover, the reaction between the melt and crucible materials can modify the composition of the melt, which typically results in growth failure, or contaminates the crystals. Thus one principle in designing a flux growth is to select suitable flux and crucible materials thus to avoid any reaction between them. In this paper, we review two cases of flux growthmore » in which the reaction between flux and Al2O3 crucible tunes the oxygen content in the melt and helps the crystallization of desired compositions. For the case of La5Pb3O, the Al2O3 crucible oxidizes La to form a passivating La2O3 layer which not only prevents further oxidization of La in the melt but also provides [O] to the melt. Finally, in the case of La0.4Na0.6Fe2As2, it is believed that the Al2O3 crucible reacts with NaAsO2 and the reaction consumes oxygen in the melt thus maintaining an oxygen-free environment.« less

  5. Reactor vessel support system

    DOE Patents [OSTI]

    Golden, Martin P.; Holley, John C.

    1982-01-01

    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.

  6. Nuclear reactor overflow line

    DOE Patents [OSTI]

    Severson, Wayne J.

    1976-01-01

    The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

  7. ISOTOPE SEPARATING APPARATUS CONTROL

    DOE Patents [OSTI]

    Barnes, S.W.

    1959-08-25

    An improved isotope separating apparatus of the electromagnetic type, commonly referred to as a calutron, is described. Improvements in detecting and maintaining optimum position and focus of the ion beam are given. The calutron collector is provided with an additional electrode insulated from and positioned between the collecting pockets. The ion beams are properly positioned and focused until the deionizing current which flows from ground to this additional electrode ts a minimum.

  8. Spinning fluids reactor

    DOE Patents [OSTI]

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

    2012-11-20

    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. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

    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.

  10. Reactor water cleanup system

    DOE Patents [OSTI]

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

    1994-12-20

    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.

  11. FORIG: a modification of the ORIGEN2 isotope-generation and depletion code for fusion problems

    SciTech Connect (OSTI)

    Blink, J.A.

    1982-03-03

    This report describes how to use the FORIG computer code to solve isotope-generation and depletion problems in fusion and fission reactors. FORIG is an adaptation of ORIGEN2 to run on a Cray-1 computer, and to accept more extensive activation cross sections.

  12. Mixed oxide fuels testing in the advanced test reactor to support plutonium disposition

    SciTech Connect (OSTI)

    Ryskamp, J.M.; Sterbentz, J.W.; Chang, G.S.

    1995-09-01

    An intense worldwide effort is now under way to find means of reducing the stockpile of weapons-grade plutonium. One of the most attractive solutions would be to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PUO{sub 2}) mixed with urania (UO{sub 2}). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification, (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania, (3) The effects of WGPu isotopic composition, (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight, (5) The effects of americium and gallium in WGPu, (6) Fission gas release from MOX fuel pellets made from WGPu, (7) Fuel/cladding gap closure, (8) The effects of power cycling and off-normal events on fuel integrity, (9) Development of radial distributions of burnup and fission products, (10) Power spiking near the interfaces of MOX and urania fuel assemblies, and (11) Fuel performance code validation. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified.

  13. Laser isotope separation of erbium and other isotopes

    DOE Patents [OSTI]

    Haynam, Christopher A.; Worden, Earl F.

    1995-01-01

    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.

  14. Laser isotope separation of erbium and other isotopes

    DOE Patents [OSTI]

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

    1995-08-22

    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.

  15. Tritium Formation and Mitigation in High-Temperature Reactors

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots

    2012-10-01

    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

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

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

    2013-03-01

    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

  17. Fast Flux Test Facility (FFTF) standby plan

    SciTech Connect (OSTI)

    Hulvey, R.K.

    1997-03-06

    The FFTF Standby Plan, Revision 0, provides changes to the major elements and project baselines to maintain the FFTF plant in a standby condition and to continue washing sodium from irradiated reactor fuel. The Plan is consistent with the Memorandum of Decision approved by the Secretary of Energy on January 17, 1997, which directed that FFTF be maintained in a standby condition to permit the Department to make a decision on whether the facility should play a future role in the Department of Energy`s dual track tritium production strategy. This decision would be made in parallel with the intended December 1998 decision on the selection of the primary, long- term source of tritium. This also allows the Department to review the economic and technical feasibility of using the FFTF to produce isotopes for the medical community. Formal direction has been received from DOE-RL and Fluor 2020 Daniel Hanford to implement the FFTF standby decision. The objective of the Plan is maintain the condition of the FFTF systems, equipment and personnel to preserve the option for plant restart within three and one-half years of a decision to restart, while continuing deactivation work which is consistent with the standby mode.

  18. Stable isotope research pool inventory

    SciTech Connect (OSTI)

    Not Available

    1988-02-01

    This report contains a listing of electromagnetically separated stable isotopes which are available at the Oak Ridge National Laboratory for distribution for nondestructive research use on a loan basis. This inventory includes all samples of stable isotopes in the Research Materials Collection and does not designate whether a sample is out on loan or is in reprocessing. For some of the high-abundance, naturally occurring isotopes, larger amounts can be made available; for example, Ca-40 and Fe-56.

  19. Stable isotope research pool inventory

    SciTech Connect (OSTI)

    Not Available

    1982-01-01

    This report contains a listing of electromagnetically separated stable isotopes which are available for distribution within the United States for nondestructive research use from the Oak Ridge National Laboratory on a loan basis. This inventory includes all samples of stable isotopes in the Material Research Collection and does not designate whether a sample is out on loan or in reprocessing. For some of the high abundance naturally occurring isotopes, larger amounts can be made available; for example, Ca-40 and Fe-56.

  20. Stable isotope research pool inventory

    SciTech Connect (OSTI)

    Not Available

    1984-03-01

    This report contains a listing of electromagnetically separated stable isotopes which are available at the Oak Ridge National Laboratory for distribution for nondestructive research use on a loan basis. This inventory includes all samples of stable isotopes in the Research Materials Collection and does not designate whether a sample is out on loan or is in reprocessing. For some of the high abundance naturally occurring isotopes, larger amounts can be made available; for example, Ca-40 and Fe-56.

  1. Antineutrino analysis for continuous monitoring of nuclear reactors: Sensitivity study

    SciTech Connect (OSTI)

    Stewart, Christopher; Erickson, Anna

    2015-10-28

    This paper explores the various contributors to uncertainty on predictions of the antineutrino source term which is used for reactor antineutrino experiments and is proposed as a safeguard mechanism for future reactor installations. The errors introduced during simulation of the reactor burnup cycle from variation in nuclear reaction cross sections, operating power, and other factors are combined with those from experimental and predicted antineutrino yields, resulting from fissions, evaluated, and compared. The most significant contributor to uncertainty on the reactor antineutrino source term when the reactor was modeled in 3D fidelity with assembly-level heterogeneity was found to be the uncertainty on the antineutrino yields. Using the reactor simulation uncertainty data, the dedicated observation of a rigorously modeled small, fast reactor by a few-ton near-field detector was estimated to offer reduction of uncertainty on antineutrino yields in the 3.0–6.5 MeV range to a few percent for the primary power-producing fuel isotopes, even with zero prior knowledge of the yields.

  2. Isotopically labeled compositions and method

    DOE Patents [OSTI]

    Schmidt, Jurgen G.; Kimball, David B.; Alvarez, Marc A.; Williams, Robert F.; Martinez, Rudolfo A.

    2011-07-12

    Compounds having stable isotopes .sup.13C and/or .sup.2H were synthesized from precursor compositions having solid phase supports or affinity tags.

  3. Korea Research Reactor -1 & 2 Decommissioning Project in Korea

    SciTech Connect (OSTI)

    Park, S. K.; Chung, U. S.; Jung, K. J.; Park, J. H.

    2003-02-24

    Korea Research Reactor 1 (KRR-1), the first research reactor in Korea, has been operated since 1962, and the second one, Korea Research Reactor 2 (KRR-2) since 1972. The operation of both of them was phased out in 1995 due to their lifetime and operation of the new and more powerful research reactor, HANARO (High-flux Advanced Neutron Application Reactor; 30MW). Both are TRIGA Pool type reactors in which the cores are small self-contained units sitting in tanks filled with cooling water. The KRR-1 is a TRIGA Mark II, which could operate at a level of up to 250 kW. The second one, the KRR-2 is a TRIGA Mark III, which could operate at a level of up 2,000 kW. The decontamination and decommissioning (D & D) project of these two research reactors, the first D & D project in Korea, was started in January 1997 and will be completed to stage 3 by 2008. The aim of this decommissioning program is to decommission the KRR-1 & 2 reactors and to decontaminate the residual building structure s and the site to release them as unrestricted areas. KAERI (Korea Atomic Energy Research Institute) submitted the decommissioning plan and the environmental impact assessment reports to the Ministry of Science and Technology (MOST) for the license in December 1998, and was approved in November 2000.

  4. ARM - Measurement - Soil moisture flux

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

    moisture flux ARM Data Discovery Browse Data Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Soil moisture flux A quantity measured according to the formula B = {lambda}(dq/dz), where {lambda} is the conductivity of the soil that the moisture is moving through. Categories Surface Properties Instruments The above measurement is considered scientifically relevant for the following instruments. Refer to the datastream (netcdf) file

  5. BWX TYmes, Vol 2, No. 9, March 14, 2002

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

    The medical community also relies heavily on Y-12's recovered materials to operate isotope research reactors, such as High Flux Isotope Reactor in Oak Ridge. There are more than ...

  6. Design of a Modular E-Core Flux Concentrating Axial Flux Machine...

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

    Design of a Modular E-Core Flux Concentrating Axial Flux Machine Preprint Tausif Husain, 1 ... Design of a Modular E-Core Flux Concentrating Axial Flux Machine Tausif Husain (1) Yilmaz ...

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

    SciTech Connect (OSTI)

    J.W. Davis

    1996-08-29

    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.

  8. Advanced isotope separation

    SciTech Connect (OSTI)

    Not Available

    1982-05-04

    The Study Group briefly reviewed the technical status of the three Advanced Isotope Separation (AIS) processes. It also reviewed the evaluation work that has been carried out by DOE's Process Evaluation Board (PEB) and the Union Carbide Corporation-Nuclear Division (UCCND). The Study Group briefly reviewed a recent draft assessment made for DOE staff of the nonproliferation implications of the AIS technologies. The staff also very briefly summarized the status of GCEP and Advanced Centrifuge development. The Study Group concluded that: (1) there has not been sufficient progress to provide a firm scientific, technical or economic basis on which to select one of the three competing AIS processes for full-scale engineering development at this time; and (2) however, should budgetary restraints or other factors force such a selection, we believe that the evaluation process that is being carried out by the PEB provides the best basis available for making a decision. The Study Group recommended that: (1) any decisions on AIS processes should include a comparison with gas centrifuge processes, and should not be made independently from the plutonium isotope program; (2) in evaluating the various enrichment processes, all applicable costs (including R and D and sales overhead) and an appropriate discounting approach should be included in order to make comparisons on a private industry basis; (3) if the three AIS programs continue with limited resources, the work should be reoriented to focus only on the most pressing technical problems; and (4) if a decision is made to develop the Atomic Vapor Laser Isotope Separation process, the solid collector option should be pursued in parallel to alleviate the potential program impact of liquid collector thermal control problems.

  9. FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

    DOE Patents [OSTI]

    Flint, O.

    1961-01-10

    Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

  10. Flux growth utilizing the reaction between flux and crucible

    SciTech Connect (OSTI)

    Yan, J. -Q.

    2015-01-22

    Flux growth involves dissolving the components of the target compound in an appropriate flux at high temperatures and then crystallizing under supersaturation controlled by cooling or evaporating the flux. A refractory crucible is generally used to contain the high temperature melt. Moreover, the reaction between the melt and crucible materials can modify the composition of the melt, which typically results in growth failure, or contaminates the crystals. Thus one principle in designing a flux growth is to select suitable flux and crucible materials thus to avoid any reaction between them. In this paper, we review two cases of flux growth in which the reaction between flux and Al2O3 crucible tunes the oxygen content in the melt and helps the crystallization of desired compositions. For the case of La5Pb3O, the Al2O3 crucible oxidizes La to form a passivating La2O3 layer which not only prevents further oxidization of La in the melt but also provides [O] to the melt. Finally, in the case of La0.4Na0.6Fe2As2, it is believed that the Al2O3 crucible reacts with NaAsO2 and the reaction consumes oxygen in the melt thus maintaining an oxygen-free environment.

  11. ISOTOPE FRACTIONATION PROCESS

    DOE Patents [OSTI]

    Clewett, G.H.; Lee, DeW.A.

    1958-05-20

    A new method is described for isotopic enrichment of uranium. It has been found that when an aqueous acidic solution of ionic tetravalent uraniunn is contacted with chelate complexed tetravalent uranium, the U/sup 238/ preferentially concentrates in the complexed phase while U/sup 235/ concentrates in the ionic phase. The effect is enhanced when the chelate compound is water insoluble and is dissolved in a water-immiscible organic solvent. Cupferron is one of a number of sultable complexing agents, and chloroform is a suitable organic solvent.

  12. Benchmarking transition costs for the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Hulvey, R.K.

    1996-12-31

    The Fast Flux Test Facility (FFTF) is a government-owned, 400-MW(thermal), sodium-cooled test reactor operated by Westinghouse Hanford Company. The reactor is shut down and is undergoing a transition to a long-term surveillance and maintenance state. The mission strategy for the FFTF transition project is to place the FFTF in a radiologically and industrially safe condition, completing the transition phase activities as soon as possible to drive down the current annual surveillance and maintenance costs from approximately $26 million/yr to roughly $1.5 million/yr. The effort to establish the shutdown and transition costs for this 7-yr, $260 million activity is a first of a kind for the U.S. Department of Energy (DOE).

  13. NEUTRONIC REACTOR SYSTEM

    DOE Patents [OSTI]

    Goett, J.J.

    1961-01-24

    A system is described which includes a neutronic reactor containing a dispersion of fissionable material in a liquid moderator as fuel and a conveyor to which a portion of the dispersion may be passed and wherein the self heat of the slurry evaporates the moderator. Means are provided for condensing the liquid moderator and returning it to the reactor and for conveying the dried fissionable material away from the reactor.

  14. Improved vortex reactor system

    DOE Patents [OSTI]

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

    1995-05-09

    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.

  15. Small Modular Reactors (SMRs) | Department of Energy

    Energy Savers [EERE]

    Reactor Technologies Small Modular Reactors (SMRs) Small Modular Reactors (SMRs) ... to the NRC by late-2016 Complete reactor module final design by mid-2019 For more ...

  16. Knowledge Preservation at the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Wootan, David W.; Omberg, Ronald P.

    2011-12-30

    One of the goals of the Department of Energy's Office of Nuclear Energy Fuel Cycle Research and Development Program (FCRD) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMRs) that could support the development of an environmentally and economically sound nuclear fuel cycle. The Fast Flux Test Facility (FFTF) is the most recent LMR to operate in the United States, from 1982 to 1992, and was designed as a fully instrumented test reactor with on-line, real time test control and performance monitoring of components and tests installed in the reactor. Knowledge preservation at the FFTF is focused on the areas of design, construction, and startup of the reactor, as well as on preserving information obtained from 10 years of successful operating history and extensive irradiation testing of fuels and materials. In order to ensure protection of information at risk, the program to date has sequestered reports, files, tapes, and drawings to allow for secure retrieval. A disciplined and orderly approach has been developed to respond to client's requests for documents and data in order to minimize the search effort and ensure that future requests for this information can be readily accommodated.

  17. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part II: Prismatic Reactor Cross Section Generation

    SciTech Connect (OSTI)

    Vincent Descotes

    2011-03-01

    The deep-burn prismatic high temperature reactor is made up of an annular core loaded with transuranic isotopes and surrounded in the center and in the periphery by reflector blocks in graphite. This disposition creates challenges for the neutronics compared to usual light water reactor calculation schemes. The longer mean free path of neutrons in graphite affects the neutron spectrum deep inside the blocks located next to the reflector. The neutron thermalisation in the graphite leads to two characteristic fission peaks at the inner and outer interfaces as a result of the increased thermal flux seen in those assemblies. Spectral changes are seen at least on half of the fuel blocks adjacent to the reflector. This spectral effect of the reflector may prevent us from successfully using the two step scheme -lattice then core calculation- typically used for light water reactors. We have been studying the core without control mechanisms to provide input for the development of a complete calculation scheme. To correct the spectrum at the lattice level, we have tried to generate cross-sections from supercell calculations at the lattice level, thus taking into account part of the graphite surrounding the blocks of interest for generating the homogenised cross-sections for the full-core calculation. This one has been done with 2 to 295 groups to assess if increasing the number of groups leads to more accurate results. A comparison with a classical single block model has been done. Both paths were compared to a reference calculation done with MCNP. It is concluded that the agreement with MCNP is better with supercells, but that the single block model remains quite close if enough groups are kept for the core calculation. 26 groups seems to be a good compromise between time and accu- racy. However, some trials with depletion have shown huge variations of the isotopic composition across a block next to the reflector. It may imply that at least an in- core depletion for the

  18. Tokamak reactor first wall

    DOE Patents [OSTI]

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

    1984-11-20

    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.

  19. Small Modular Reactors - SRSCRO

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

    smr Small Modular Reactors The Savannah River National Laboratory (SRNL) has announced several partnerships to bring refrigerator-sized modular nuclear reactors, known as Small Modular Reactors or SMRs, to the Savannah River Site facility and jump start development of the U.S. Energy Freedom CenterTM. Currently, all large commercial power reactors in the United States and most in the rest of the world are based on "light water" designs - that is, they use uranium fuel and ordinary

  20. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, Juhani

    1996-01-01

    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. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, J.

    1996-03-19

    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.

  2. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOE Patents [OSTI]

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  3. Method of fission heat flux determination from experimental data

    DOE Patents [OSTI]

    Paxton, Frank A.

    1999-01-01

    A method is provided for determining the fission heat flux of a prime specimen inserted into a specimen of a test reactor. A pair of thermocouple test specimens are positioned at the same level in the holder and a determination is made of various experimental data including the temperature of the thermocouple test specimens, the temperature of bulk water channels located in the test holder, the gamma scan count ratios for the thermocouple test specimens and the prime specimen, and the thicknesses of the outer clads, the fuel fillers, and the backclad of the thermocouple test specimen. Using this experimental data, the absolute value of the fission heat flux for the thermocouple test specimens and prime specimen can be calculated.

  4. Isotope Program Transportation | Department of Energy

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

    Isotope Program Transportation Isotope Program Transportation Isotope Program Transportation (894.11 KB) More Documents & Publications Nuclear Fuel Storage and Transportation Planning Project Overview Section 180(c) Ad Hoc Working Group DOE Office of Nuclear Energy

  5. Gas tagging and cover gas combination for nuclear reactor

    DOE Patents [OSTI]

    Gross, Kenny C.; Laug, Matthew T.

    1985-01-01

    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.

  6. AmeriFlux US-Sta Saratoga

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

    Ewers, Brent [University of Wyoming; Pendall, Elise [University of Wyoming

    2016-01-01

    This is the AmeriFlux version of the carbon flux data for the site US-Sta Saratoga. Site Description - Sagebrush steppe ecosystem

  7. AmeriFlux US-Wdn Walden

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

    Ewers, Brent [University of Wyoming; Pendall, Elise [University of Wyoming

    2016-01-01

    This is the AmeriFlux version of the carbon flux data for the site US-Wdn Walden. Site Description - Sagebrush steppe ecosystem

  8. Advanced Models of LWR Pressure Vessel Embrittlement for Low Flux-HighFluence Conditions

    SciTech Connect (OSTI)

    Odette, G. Robert; Yamamoto, Takuya

    2013-06-17

    Neutron embrittlement of reactor pressure vessels (RPVs) is an unresolved issue for light water reactor life extension, especially since transition temperature shifts (TTS) must be predicted for high 80-year fluence levels up to approximately 1,020 n/cm{sup 2}, far beyond the current surveillance database. Unfortunately, TTS may accelerate at high fluence, and may be further amplified by the formation of late blooming phases that result in severe embrittlement even in low-copper (Cu) steels. Embrittlement by this mechanism is a potentially significant degradation phenomenon that is not predicted by current regulatory models. This project will focus on accurately predicting transition temperature shifts at high fluence using advanced physically based, empirically validated and calibrated models. A major challenge is to develop models that can adjust test reactor data to account for flux effects. Since transition temperature shifts depend on synergistic combinations of many variables, flux-effects cannot be treated in isolation. The best current models systematically and significantly under-predict transition temperature at high fluence, although predominantly for irradiations at much higher flux than actual RPV service. This project will integrate surveillance, test reactor and mechanism data with advanced models to address a number of outstanding RPV embrittlement issues. The effort will include developing new databases and preliminary models of flux effects for irradiation conditions ranging from very low (e.g., boiling water reactor) to high (e.g., accelerated test reactor). The team will also develop a database and physical models to help predict the conditions for the formation of Mn-Ni-Si late blooming phases and to guide future efforts to fully resolve this issue. Researchers will carry out other tasks on a best-effort basis, including prediction of transition temperature shift attenuation through the vessel wall, remediation of embrittlement by annealing

  9. Phonon coherence in isotopic silicon superlattices

    SciTech Connect (OSTI)

    Frieling, R.; Radek, M.; Eon, S.; Bracht, H.; Wolf, D. E.

    2014-09-29

    Recent experimental and theoretical investigations have confirmed that a reduction in thermal conductivity of silicon is achieved by isotopic silicon superlattices. In the present study, non-equilibrium molecular dynamics simulations are performed to identify the isotope doping and isotope layer ordering with minimum thermal conductivity. Furthermore, the impact of isotopic intermixing at the superlattice interfaces on phonon transport is investigated. Our results reveal that the coherence of phonons in isotopic Si superlattices is prevented if interfacial mixing of isotopes is considered.

  10. Code System for Reactor Physics and Fuel Cycle Simulation.

    Energy Science and Technology Software Center (OSTI)

    1999-04-21

    Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterativemore » processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.« less

  11. Background radiation measurements at high power research reactors

    SciTech Connect (OSTI)

    Ashenfelter, J.; Yeh, M.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Davee, D.; Dean, D.; Deichert, G.; Dolinski, M. J.; Dolph, J.; Dwyer, D. A.; Fan, S.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Green, M.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Jaffe, D. E.; Kettell, S.; Langford, T. J.; Littlejohn, B. R.; Martinez, D.; McKeown, R. D.; Morrell, S.; Mueller, P. E.; Mumm, H. P.; Napolitano, J.; Norcini, D.; Pushin, D.; Romero, E.; Rosero, R.; Saldana, L.; Seilhan, B. S.; Sharma, R.; Stemen, N. T.; Surukuchi, P. T.; Thompson, S. J.; Varner, R. L.; Wang, W.; Watson, S. M.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yen, Y. -R.; Zhang, C.; Zhang, X.

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  12. Background radiation measurements at high power research reactors

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

    Ashenfelter, J.; Yeh, M.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; et al

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the backgroundmore » fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.« less

  13. Method of separating boron isotopes

    DOE Patents [OSTI]

    Jensen, Reed J.; Thorne, James M.; Cluff, Coran L.; Hayes, John K.

    1984-01-01

    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.

  14. Method of separating boron isotopes

    DOE Patents [OSTI]

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

    1981-01-23

    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/.

  15. Stable isotope research pool inventory

    SciTech Connect (OSTI)

    Not Available

    1980-12-01

    This report contains a listing of electromagnetically separated stable isotopes which are available for distribution within the United States for non-destructive research use from the Oak Ridge National Laboratory on a loan basis. This inventory includes all samples of stable isotopes in the Materials Research Collection and does not designate whether a sample is out on loan or in reprocessing.

  16. Order 13287, Preserve America

    Energy Savers [EERE]

    ... In July 2014, the High Flux Isotope Reactor has been approved as an ANS Nuclear Historic ... Many of these stable isotopes were used to create medical isotopes that are still used in ...

  17. Heavy Isotopes Lead Materials Management Organization (LMMO)...

    Office of Scientific and Technical Information (OSTI)

    Heavy Isotopes Lead Materials Management Organization (LMMO) Update Citation Details In-Document Search Title: Heavy Isotopes Lead Materials Management Organization (LMMO) Update ...

  18. Isotopic Trends in Production of Superheavies

    SciTech Connect (OSTI)

    Antonenko, N.V.; Adamian, G.G.; Zubov, A.S.; Scheid, W.

    2005-11-21

    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.

  19. Isotope separation by laser means

    DOE Patents [OSTI]

    Robinson, C. Paul; Jensen, Reed J.; Cotter, Theodore P.; Greiner, Norman R.; Boyer, Keith

    1982-06-15

    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.

  20. 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...

  1. Critical Heat Flux Phenomena at HighPressure & Low Mass Fluxes: NEUP Final Report Part I: Experiments

    SciTech Connect (OSTI)

    Corradini, Michael; Wu, Qiao

    2015-04-30

    This report is a preliminary document presenting an overview of the Critical Heat Flux (CHF) phenomenon, the High Pressure Critical Heat Flux facility (HPCHF), preliminary CHF data acquired, and the future direction of the research. The HPCHF facility has been designed and built to study CHF at high pressure and low mass flux ranges in a rod bundle prototypical of conceptual Small Modular Reactor (SMR) designs. The rod bundle is comprised of four electrically heated rods in a 2x2 square rod bundle with a prototypic chopped-cosine axial power profile and equipped with thermocouples at various axial and circumferential positions embedded in each rod for CHF detection. Experimental test parameters for CHF detection range from pressures of ~80 – 160 bar, mass fluxes of ~400 – 1500 kg/m2s, and inlet water subcooling from ~30 – 70°C. The preliminary data base established will be further extended in the future along with comparisons to existing CHF correlations, models, etc. whose application ranges may be applicable to the conditions of SMRs.

  2. NEUTRONIC REACTOR BURIAL ASSEMBLY

    DOE Patents [OSTI]

    Treshow, M.

    1961-05-01

    A burial assembly is shown whereby an entire reactor core may be encased with lead shielding, withdrawn from the reactor site and buried. This is made possible by a five-piece interlocking arrangement that may be easily put together by remote control with no aligning of bolt holes or other such close adjustments being necessary.

  3. NEUTRONIC REACTOR SYSTEM

    DOE Patents [OSTI]

    Treshow, M.

    1959-02-10

    A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

  4. Light water reactor program

    SciTech Connect (OSTI)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  5. REFLECTOR FOR NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Fraas, A.P.

    1963-08-01

    A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

  6. 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

    2012-06-01

    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.

  7. Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements

    SciTech Connect (OSTI)

    J. D. Bess; T. L. Maddock; M. A. Marshall

    2011-09-01

    The Neutron Radiography (NRAD) reactor is a 250-kW TRIGA-(Training, Research, Isotope Production, General Atomics)-conversion-type reactor at the Idaho National Laboratory; it is primarily used for neutron radiography analysis of irradiated and unirradiated fuels and materials. The NRAD reactor was converted from HEU to LEU fuel with 60 fuel elements and brought critical on March 31, 2010. This configuration of the NRAD reactor has been evaluated as an acceptable benchmark experiment and is available in the 2011 editions of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Significant effort went into precisely characterizing all aspects of the reactor core dimensions and material properties; detailed analyses of reactor parameters minimized experimental uncertainties. The largest contributors to the total benchmark uncertainty were the 234U, 236U, Er, and Hf content in the fuel; the manganese content in the stainless steel cladding; and the unknown level of water saturation in the graphite reflector blocks. A simplified benchmark model of the NRAD reactor was prepared with a keff of 1.0012 {+-} 0.0029 (1s). Monte Carlo calculations with MCNP5 and KENO-VI and various neutron cross section libraries were performed and compared with the benchmark eigenvalue for the 60-fuel-element core configuration; all calculated eigenvalues are between 0.3 and 0.8% greater than the benchmark value. Benchmark evaluations of the NRAD reactor are beneficial in understanding biases and uncertainties affecting criticality safety analyses of storage, handling, or transportation applications with LEU-Er-Zr-H fuel.

  8. Neutron Scattering Facilities | U.S. DOE Office of Science (SC...

    Office of Science (SC) Website

    High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory The HFIR facility is the United States' highest flux reactor-based neutron source, and is a major neutron ...

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

    SciTech Connect (OSTI)

    Forsberg, C.

    2013-07-01

    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)

  10. Some Aspects of Reactor Theory

    DOE R&D Accomplishments [OSTI]

    Weinberg, Alvin M.

    1952-10-10

    Some general remarks are made on reactor theory, particularly the asymptotic theory and multigroup methods. Unsolved reactor problems are also briefly discussed. (B.J.H.)

  11. Reactor Materials | Department of Energy

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

    reactor materials crosscut effort will enable the development of innovative and ... Research into specific degradation modes or material needs unique to a particular reactor ...

  12. Container for hydrogen isotopes

    DOE Patents [OSTI]

    Solomon, David E.

    1977-01-01

    A container for the storage, shipping and dispensing of hydrogen isotopes such as hydrogen, deuterium, tritium, or mixtures of the same which has compactness, which is safe against fracture or accident, and which is reusable. The container consists of an outer housing with suitable inlet and outlet openings and electrical feed elements, the housing containing an activated sorber material in the form, for example, of titanium sponge or an activated zirconium aluminate cartridge. The gas to be stored is introduced into the chamber under conditions of heat and vacuum and will be retained in the sorber material. Subsequently, it may be released by heating the unit to drive off the stored gas at desired rates.

  13. High flux solar energy transformation

    DOE Patents [OSTI]

    Winston, R.; Gleckman, P.L.; O'Gallagher, J.J.

    1991-04-09

    Disclosed are multi-stage systems for high flux transformation of solar energy allowing for uniform solar intensification by a factor of 60,000 suns or more. Preferred systems employ a focusing mirror as a primary concentrative device and a non-imaging concentrator as a secondary concentrative device with concentrative capacities of primary and secondary stages selected to provide for net solar flux intensification of greater than 2000 over 95 percent of the concentration area. Systems of the invention are readily applied as energy sources for laser pumping and in other photothermal energy utilization processes. 7 figures.

  14. ARM - Measurement - Soil heat flux

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

    heat flux ARM Data Discovery Browse Data Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Soil heat flux A quantity measured according to the formula B = {lambda}(dT/dz), where {lambda} is the conductivity of the soil that the heat is moving through. Categories Surface Properties Instruments The above measurement is considered scientifically relevant for the following instruments. Refer to the datastream (netcdf) file headers of each

  15. High flux solar energy transformation

    DOE Patents [OSTI]

    Winston, Roland; Gleckman, Philip L.; O'Gallagher, Joseph J.

    1991-04-09

    Disclosed are multi-stage systems for high flux transformation of solar energy allowing for uniform solar intensification by a factor of 60,000 suns or more. Preferred systems employ a focusing mirror as a primary concentrative device and a non-imaging concentrator as a secondary concentrative device with concentrative capacities of primary and secondary stages selected to provide for net solar flux intensification of greater than 2000 over 95 percent of the concentration area. Systems of the invention are readily applied as energy sources for laser pumping and in other photothermal energy utilization processes.

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

    SciTech Connect (OSTI)

    John D. Bess

    2014-03-01

    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.

  17. MANTA. An Integral Reactor Physics Experiment to Infer the Neutron Capture Cross Sections of Actinides and Fission Products in Fast and Epithermal Spectra

    SciTech Connect (OSTI)

    Youinou, Gilles Jean-Michel

    2015-10-01

    Neutron cross-sections characterize the way neutrons interact with matter. They are essential to most nuclear engineering projects and, even though theoretical progress has been made as far as the predictability of neutron cross-section models, measurements are still indispensable to meet tight design requirements for reduced uncertainties. Within the field of fission reactor technology, one can identify the following specializations that rely on the availability of accurate neutron cross-sections: (1) fission reactor design, (2) nuclear fuel cycles, (3) nuclear safety, (4) nuclear safeguards, (5) reactor monitoring and neutron fluence determination and (6) waste disposal and transmutation. In particular, the assessment of advanced fuel cycles requires an extensive knowledge of transuranics cross sections. Plutonium isotopes, but also americium, curium and up to californium isotope data are required with a small uncertainty in order to optimize significant features of the fuel cycle that have an impact on feasibility studies (e.g. neutron doses at fuel fabrication, decay heat in a repository, etc.). Different techniques are available to determine neutron cross sections experimentally, with the common denominator that a source of neutrons is necessary. It can either come from an accelerator that produces neutrons as a result of interactions between charged particles and a target, or it can come from a nuclear reactor. When the measurements are performed with an accelerator, they are referred to as differential since the analysis of the data provides the cross-sections for different discrete energies, i.e. σ(Ei), and for the diffusion cross sections for different discrete angles. Another approach is to irradiate a very pure sample in a test reactor such as the Advanced Test Reactor (ATR) at INL and, after a given time, determine the amount of the different transmutation products. The precise characterization of the nuclide densities before and after

  18. Preserving Physics Knowledge at the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Wootan, David W.; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.; Nelson, Joseph V.; Polzin, David L.

    2011-11-01

    One of the goals of the Department of Energy’s Office of Nuclear Energy, initiated under the Fuel Cycle Research and Development Program (FCRD) and continued under the Advanced Reactor Concepts Program (ARC) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMRs) that could support the development of an environmentally and economically sound nuclear fuel cycle. The Fast Flux Test Facility (FFTF) is the most recent LMR to operate in the United States, from 1982 to 1992, and was designed as a fully instrumented test reactor with on-line, real time test control and performance monitoring of components and tests installed in the reactor. The 10 years of operation of the FFTF provided a very useful framework for testing the advances in LMR safety technology based on passive safety features that may be of increased importance to new designs after the events at Fukushima. Knowledge preservation at the FFTF is focused on the areas of design, construction, and startup of the reactor, as well as on preserving information obtained from 10 years of successful operating history and extensive irradiation testing of fuels and materials. In order to ensure protection of information at risk, the program to date has sequestered reports, files, tapes, and drawings to allow for secure retrieval. A disciplined and orderly approach has been developed to respond to client’s requests for documents and data in order to minimize the search effort and ensure that future requests for this information can be readily accommodated.

  19. Knowledge Management at the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Wootan, David W.; Omberg, Ronald P.

    2013-06-01

    One of the goals of the Department of Energy’s Office of Nuclear Energy, initiated under the Fuel Cycle Research and Development Program (FCRD) and continued under the Advanced Reactor Concepts Program (ARC) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMRs) that could support the development of an environmentally and economically sound nuclear fuel cycle. The Fast Flux Test Facility (FFTF) is the most recent LMR to operate in the United States, from 1982 to 1992, and was designed as a fully instrumented test reactor with on-line, real time test control and performance monitoring of components and tests installed in the reactor. The 10 years of operation of the FFTF provided a very useful framework for testing the advances in LMR safety technology based on passive safety features that may be of increased importance to new designs after the events at Fukushima. Knowledge preservation at the FFTF is focused on the areas of design, construction, and startup of the reactor, as well as on preserving information obtained from 10 years of successful operating history and extensive irradiation testing of fuels and materials. In order to ensure protection of information at risk, the program to date has sequestered reports, files, tapes, and drawings to allow for secure retrieval. The FFTF knowledge management program includes a disciplined and orderly approach to respond to client’s requests for documents and data in order to minimize the search effort and ensure that future requests for this information can be readily accommodated.

  20. TEST-HOLE CONSTRUCTION FOR A NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Ohlinger, L.A.; Seitz, F.; Young, G.J.

    1959-02-17

    Test-hole construction is described for a reactor which provides safe and ready access to the neutron flux region for specimen materials which are to be irradiated therein. An elongated tubular thimble adapted to be inserted in the access hole through the wall of the reactor is constructed of aluminum and is provided with a plurality of holes parallel to the axis of the thimble for conveying the test specimens into position for irradiation, and a conduit for the circulation of coolant. A laminated shield formed of alternate layers of steel and pressed wood fiber is disposed lengthwise of the thimble near the outer end thereof.

  1. Stable isotope research pool inventory

    SciTech Connect (OSTI)

    Not Available

    1981-01-01

    This report contains a listing of electromagnetically separated stable isotopes which are available for distribution within the United States for non-destructive research use from the Oak Ridge National Laboratory on a loan basis. This inventory includes all samples of stable isotopes in the Materials Research Collection and does not designate whether a sample is out on loan or in reprocessing. For some of the high abundance naturally occurring isotopes, larger amounts can be made available; for example, Ca-40 and Fe-56.

  2. Stable isotope research pool inventory

    SciTech Connect (OSTI)

    Not Available

    1983-03-01

    This report contains a listing of electromagnetically separated stable isotopes which are available for distribution within the United States for non-destructive research use from the Oak Ridge National Laboratory on a loan basis. This inventory includes all samples of stable isotopes in the Materials Research Collection and does not designate whether a sample is out on loan or in reprocessing. For some of the high abundance naturally occurring isotopes, larger amounts can be made available; for example, Ca-40 and Fe-56.

  3. Compelling Research Opportunities using Isotopes

    SciTech Connect (OSTI)

    2009-04-23

    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

  4. Nuclear reactor control column

    DOE Patents [OSTI]

    Bachovchin, Dennis M.

    1982-01-01

    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.

  5. Slurry reactor design studies

    SciTech Connect (OSTI)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. ); Akgerman, A. ); Smith, J.M. )

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  6. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Edler, S. K.

    1981-07-01

    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.

  7. Method and apparatus for removing iodine from a nuclear reactor coolant

    DOE Patents [OSTI]

    Cooper, Martin H.

    1980-01-01

    A method and apparatus for removing iodine-131 and iodine-125 from a liquid sodium reactor coolant. Non-radioactive iodine is dissolved in hot liquid sodium to increase the total iodine concentration. Subsequent precipitation of the iodine in a cold trap removes both the radioactive iodine isotopes as well as the non-radioactive iodine.

  8. Superconducting flux flow digital circuits

    DOE Patents [OSTI]

    Hietala, Vincent M.; Martens, Jon S.; Zipperian, Thomas E.

    1995-01-01

    A NOR/inverter logic gate circuit and a flip flop circuit implemented with superconducting flux flow transistors (SFFTs). Both circuits comprise two SFFTs with feedback lines. They have extremely low power dissipation, very high switching speeds, and the ability to interface between Josephson junction superconductor circuits and conventional microelectronics.

  9. Superconducting flux flow digital circuits

    DOE Patents [OSTI]

    Hietala, V.M.; Martens, J.S.; Zipperian, T.E.

    1995-02-14

    A NOR/inverter logic gate circuit and a flip flop circuit implemented with superconducting flux flow transistors (SFFTs) are disclosed. Both circuits comprise two SFFTs with feedback lines. They have extremely low power dissipation, very high switching speeds, and the ability to interface between Josephson junction superconductor circuits and conventional microelectronics. 8 figs.

  10. Preparation of actinide specimens for the US/UK joint experiment in the Dounreay Prototype Fast Reactor

    SciTech Connect (OSTI)

    Quinby, T C; Adair, H L; Kobisk, E H

    1982-05-01

    A joint research program involving the United States and the United Kingdom was initiated about four years ago for the purpose of studying the fuel behavior of higher actinides using in-core irradiation in the fast reactor at Dounreay, Scotland. Simultaneously, determination of integral cross sections of a wide variety of higher actinide isotopes (physics specimens) was proposed. Coincidental neutron flux and energy spectral measurements were to be made using vanadium encapsulated dosimetry materials in the immediate region of the fuel pellets and physics samples. The higher actinide samples chosen for the fuel study were /sup 241/Am and /sup 244/Cm in the forms of Am/sub 2/O/sub 3/, Cm/sub 2/O/sub 3/, and Am/sub 6/Cm(RE)/sub 7/O/sub 21/, where (RE) represents a mixture of lanthanides. Milligram quantities of actinide oxides of /sup 248/Cm, /sup 246/Cm, /sup 244/Cm, /sup 243/Cm, /sup 243/Am, /sup 241/Am, /sup 244/Pu, /sup 242/Pu, /sup 241/Pu, /sup 240/Pu, /sup 239/Pu, /sup 238/Pu, /sup 237/Np, /sup 238/U, /sup 236/U, /sup 235/U, /sup 234/U, /sup 233/U, /sup 232/Th, /sup 230/Th, and /sup 231/Pa were encapsulated to obtain nuclear cross section and reaction rate data for these materials.

  11. Apparatus and process for separating hydrogen isotopes

    DOE Patents [OSTI]

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

    2013-06-25

    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.

  12. Method for laser induced isotope enrichment

    DOE Patents [OSTI]

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

    2004-09-07

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

  13. Alternative applications of atomic vapor laser isotope separation technology

    SciTech Connect (OSTI)

    Not Available

    1991-01-01

    This report was commissioned by the Secretary of Energy. It summarizes the main features of atomic vapor laser isotope separation (AVLIS) technology and subsystems; evaluates applications, beyond those of uranium enrichment, suggested by Lawrence Livermore National Laboratory (LLNL) and a wide range of US industries and individuals; recommends further work on several applications; recommends the provision of facilities for evaluating potential new applications; and recommends the full involvement of end users from the very beginning in the development of any application. Specifically excluded from this report is an evaluation of the main AVLIS missions, uranium enrichment and purification of plutonium for weapons. In evaluating many of the alternative applications, it became clear that industry should play a greater and earlier role in the definition and development of technologies with the Department of Energy (DOE) if the nation is to derive significant commercial benefit. Applications of AVLIS to the separation of alternate (nonuranium) isotopes were considered. The use of {sup 157}Gd as burnable poison in the nuclear fuel cycle, the use {sup 12}C for isotopically pure diamond, and the use of plutonium isotopes for several nonweapons applications are examples of commercially useful products that might be produced at a cost less than the product value. Separations of other isotopes such as the elemental constituents of semiconductors were suggested; it is recommended that proposed applications be tested by using existing supplies to establish their value before more efficient enrichment processes are developed. Some applications are clear, but their production costs are too high, the window of opportunity in the market has passed, or societal constraints (e.g., on reprocessing of reactor fuel) discourage implementation.

  14. 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.

    1980-07-01

    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.

  15. CONTROL FOR NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Lichtenberger, H.V.; Cameron, R.A.

    1959-03-31

    S>A control rod operating device in a nuclear reactor of the type in which the control rod is gradually withdrawn from the reactor to a position desired during stable operation is described. The apparatus is comprised essentially of a stop member movable in the direction of withdrawal of the control rod, a follower on the control rod engageable with the stop and means urging the follower against the stop in the direction of withdrawal. A means responsive to disengagement of the follower from the stop is provided for actuating the control rod to return to the reactor shut-down position.

  16. Microfluidic electrochemical reactors

    DOE Patents [OSTI]

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

    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.

  17. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, Ronald J.; Land, John T.; Misvel, Michael C.

    1994-01-01

    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.

  18. Nuclear reactor reflector

    DOE Patents [OSTI]

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

    1994-06-07

    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.

  19. COOLED NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Binner, C.R.; Wilkie, C.B.

    1958-03-18

    This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

  20. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Martin Peng, Y.K.M.

    1985-10-03

    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.