Browsing Center for Advanced Nuclear Energy Systems (CANES) by Title

Now showing items 21-40 of 109

    • Conceptual Reactor Physics Design of a Lead-Bismuth-Cooled Critical Actinide Burner 

      Hejzlar, Pavel; Driscoll, Michael J.; Kazimi, Mujid S. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2000-02)
      Destruction of actinides in accelerator-driven subcriticals and in stand-alone critical reactors is of contemporary interest as a means to reduce long-term high-level waste radiotoxicity. This topical report is focused ...

    • Critical Heat Flux Enhancement via Surface Modification Using Colloidal Dispersions of Nanoparticles (Nanofluids) 

      Truong, Bao H.; Hu, Lin-Wen; Buongiorno, Jacopo (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2008-06)
      Nanofluids are engineered colloidal dispersions of nanoparticles (1-100nm) in common fluids (water, refrigerants, or ethanol…). Materials used for nanoparticles include chemically stable metals (e.g., gold, silver, ...

    • Cross Section Generation Strategy for High Conversion Light Water Reactors 

      Herman, Bryan R.; Shwageraus, Eugene; Forget, Benoit; Kazimi, Mujid S. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program, 2011-06)
      High conversion water reactors (HCWR), such as the Resource-renewable Boiling Water Reactor (RBWR), are being designed with axial heterogeneity of alternating fissile and blanket zones to achieve a conversion ratio of ...

    • Deep Boreholes Attributes and Performance Requirements 

      Driscoll, Michael J.; Jensen, K. G. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program, 2010-05-01)
      This is a progress report covering work through mid-May 2010 under a Sandia-MIT contract dealing with design and siting/licensing criteria for deep borehole disposal of spent nuclear fuel or its separated constituents. It ...

    • Design and Optimization of a High Thermal Flux Research Reactor Via Kriging-Based Algorithm 

      Kempf, Stephanie A.; Hu, Lin-Wen; Forget, Benoit (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. MIT Reactor Redesign Program, 2011-06-01)
      In response to increasing demands for the services of research reactors, a 5 MW LEUfueled research reactor core is developed and optimized to provide high thermal flux within specified limits upon thermal hydraulic ...

    • Design of a Functionally Graded Composite for Service in High Temperature Lead and Lead-Bismuth Cooled Nuclear Reactors 

      Short, Michael P.; Ballinger, Ronald G. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2010-10)
      A material that resists lead-bismuth eutectic (LBE) attack and retains its strength at 700°C would be an enabling technology for LBE-cooled reactors. No single alloy currently exists that can economically meet the required ...

    • Design of a Low Enrichment, Enhanced Fast Flux Core for the MIT Research Reactor 

      Ellis, T.S.; Forget, Benoit; Kazimi, Mujid S.; Newton, T.; Pilat, Edward E. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. MIT Reactor Redesign Program, 2009-02-01)
      Worldwide, there is limited test reactor capacity to perform the required irradiation experiments on advanced fast reactor materials and fuel designs. This is particularly true in the U.S., which no longer has an operating ...

    • Design of Compact Intermediate Heat Exchangers for Gas Cooled Fast Reactors 

      Gezelius, K.; Driscoll, Michael J. ;; Hejzlar, Pavel (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2004-05)
      Two aspects of an intermediate heat exchanger (IHX) for GFR service have been investigated: (1) the intrinsic characteristics of the proposed compact printed circuit heat exchanger (PCHE); and (2) a specific design ...

    • Developing Fuel Management Capabilities Based On Coupled Monte Carlo Depletion in Support of the MIT Research Reactor Conversion 

      Romano, Paul Kollath; Newton, Thomas H., Jr.; Forget, Benoit (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. MIT Reactor Redesign Program, 2009-06-01)
      Pursuant to a 1986 NRC ruling, the MIT Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. Prior studies have shown that the MITR will be able ...

    • Development of a Bayesian Network to Monitor the Probability of Nuclear Proliferation 

      Holcombe, Robert; Golay, Michael W. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program, 2010-04)
      Nuclear Proliferation is a complex problem that has plagued national security strategists since the advent of the first nuclear weapons. As the cost to produce nuclear weapons has continued to decline and the availability ...

    • A Drop-In Concept for Deep Borehole Canister Emplacement 

      Bates, Ethan A.; Buongiorno, Jacopo; Driscoll, Michael J. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program, 2011-06)
      Disposal of high-level nuclear waste in deep boreholes drilled into crystalline bedrock (i.e., “granite”) is an interesting repository alternative of long standing. Work at MIT over the past two decades, and more recently ...

    • Effective Thermal Conductivity of Prismatic MHTGR Fuel 

      Han, J. C.; Driscoll, M. J.; Todreas, N. E. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 1989-09-30)
      The Reactor Cavity Cooling System (RCCS) is an essential passive safety feature of the Modular High Temperature Gas-Cooled Reactor (MHTGR). Its function is to assure the protection of both public safety and owner investment. ...

    • Effects of Surface Parameters on Boiling Heat Transfer Phenomena 

      Truong, Bao Hoai; Hu, Lin-wen; Buongiorno, Jacopo; McKrell, Thomas J. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2011-06)
      Nanofluids, engineered colloidal dispersions of nanoparticles in fluid, have been shown to enhance pool and flow boiling CHF. The CHF enhancement was due to nanoparticle deposited on the heater surface, which was verified ...

    • Estimate of Radiation Release from MIT Reactor with Low Enriched Uranium (LEU) Core During Maximum Hypothetical Accident 

      Plumer, Kevin E.; Newton, Thomas H., Jr.; Forget, Benoit (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. MIT Reactor Redesign Program, 2011-06-01)
      In accordance with a 1986 NRC ruling, the MIT Research Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. A component of the conversion analysis ...

    • Evaluation of the Thermal-Hydraulic Operating Limits of HEU-LEU Transition Cores for the MIT Research Reactor 

      Wan, Yunzhi; Hu, Lin-Wen (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. MIT Reactor Redesign Program, 2009-05-01)
      The MIT Research Reactor (MITR) is in the process of conducting a design study to convert from High Enrichment Uranium (HEU) fuel to Low Enrichment Uranium (LEU) fuel. The currently selected LEU fuel design contains 18 ...

    • AN EVOLUTIONARY FUEL ASSEMBLY DESIGN FOR HIGH POWER DENSITY BWRs 

      Karahan, A.; Buongiorno, Jacopo; Kazimi, Mujid S. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program, 2007-07)
      An evolutionary BWR fuel assembly design was studied as a means to increase the power density of current and future BWR cores. The new assembly concept is based on replacing four traditional assemblies and large water ...

    • Experimental Determination of Thermal Conductivity of a Lead- Bismuth, Eutectic-Filled Annulus 

      Carpenter, David M.; Kohse, Gordon E. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program, 2005-06)
      In order to obtain an accurate prediction of the thermal behavior of an annular fuel assembly (see MIT-NFC-PR-048 for a description of the rods), the thermal conduction of the region from the outside of the fuel capsule ...

    • Feasibility Investigations for Risk-Based Nuclear Safety Regulation 

      Beer, B. C.; Golay, M. W.; Apostolakis, G. E. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Systems Enhanced Performance Program, 2001-02)

    • Feasibility of Breeding in Hard Spectrum Boiling Water Reactors with Oxide and Nitride Fuels 

      Feng, Bo; Kazimi, Mujid S.; Forget, Benoit (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program, 2011-06-01)
      This study assesses the neutronic, thermal-hydraulic, and fuel performance aspects of using nitride fuel in place of oxides in Pu-based high conversion light water reactor designs. Using the higher density nitride fuel ...

    • Feasibility of Partial ZrO[subscript 2] Coatings on Outer Surface of Annular UO[subscript 2] Pellets to Control Gap Conductance 

      Feinroth, H.; Yuan, Y. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program, 2005-06)
      The viability of depositing a thin porous coating of zirconia on the outer surface of an annular UO[subscript 2] pellet is investigated experimentally. Such a layer has been proposed to buffer the contact between the fuel ...