Browsing Advanced Nuclear Power Technology Program (ANP) - Technical Reports by Issue Date

Now showing items 1-20 of 34

    • 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. ...

    • Conceptual Design of a Large, Passive Pressure-Tube Light Water Reactor 

      Hejzlar, P.; Todreas, N. E.; Driscoll, M. J. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 1994-06)
      A design for a large, passive, light water reactor has been developed. The proposed concept is a pressure tube reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated ...

    • Conceptual Design of a Large, Passive Pressure-Tube Light Water Reactor 

      Hejzlar, P.; Todreas, N. E.; Driscoll, M. J. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 1994-06-01)
      A design for a large, passive, light water reactor has been developed. The proposed concept is a pressure tube reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid Sic-coated ...

    • An Integrated Formal Approach for Developing High Quality Software for Safety-Critical Systems 

      Ouyang, Meng; Golay, Michael W. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 1995-09)
      This report presents the results of a study which devises an Integrated Formal Approach (IFA) for improving specifications of the designs of computer programs used in safety-critical systems. In this IFA, the formal ...

    • Use of Performance Monitoring to Improve Reliability of Emergency Generators Diesel 

      Dulik, J. D.; Golay, M. W. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 1997-12)
      Emergency diesel generators are one of the most important contributors to the core damage failure rate of nuclear power plants. Current required testing and maintenance procedures are excessively strict and expensive without ...

    • Risk-Informed, Performance-Based Regulatory Implications of Improved Emergency Diesel Generator Reliability 

      Utton, S.; Golay, M. W. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 1998-01)
      The Nuclear Regulatory Commission's (NRC) steady progress towards risk-informed performance-based regulation (RIPBR) prompted the practical application of this regulatory tool in order to demonstrate its potential benefits. ...

    • A Semi-Passive Containment Cooling System Conceptual Design 

      Liu, H.; Todreas, N. E.; Driscoll, M. J.; Byun, C. S.; Kim, Y. H.; e.a. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 1998-02)
      The objective of this project was to investigate a passive containment cooling system (PCCS) for the double concrete containment of the Korean Next Generation Reactor (KNGR). Two conceptual PCCS designs: the thermosyphon ...

    • 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 ...

    • Selection of Correlations and Look-Up Tables for Critical Heat Flux Prediction in the Generation IV "IRIS" Reactor 

      Romano, A.; Todreas, Neil E. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2000-06)
      In order to fulfill the goals set forth by the Generation IV International Forum, the current NERI funded research has focused on the design of a Gas-cooled Fast Reactor (GFR) operating in a Breed and Burn (B&B) fuel ...

    • Modular Pebble Bed Reactor 

      Kadak, Andrew C.; Ballinger, Ronald G.; Driscoll, Michael J.; Yip, Sidney; Wilson, David Gordon; e.a. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2000-07)
      This project is developing a fundamental conceptual design for a gas-cooled, modular, pebble bed reactor. Key technology areas associated with this design are being investigated which intend to address issues concerning ...

    • Plant Design and Cost Estimation of a Natural Circulation Lead-Bismuth Reactor with Steam Power Conversion Cycle 

      Kim, D.; Todreas, N. E.; Kazimi, Mujid S.; Driscoll, M. J. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2000-08)
      The analysis of an indirect steam power conversion system with lead-bismuth natural circulation primary system has been performed. The work of this report is focused on 1) identifying the allowable design region for the ...

    • Plant Design and Cost Estimation of a Natural Circulation Lead-Bismuth Reactor with Helium Power Conversion Cycle 

      Kim, D.; Todreas, N. E.; Kazimi, Mujid S.; Driscoll, M. J. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2000-11)
      The analysis of an indirect helium power conversion system with lead-bismuth natural circulation primary system has been performed. The work of this report is focused on 1) identifying the allowable design region for the ...

    • Conceptual Design of a Lead-Bismuth Cooled Fast Reactor with In-Vessel Direct-Contact Steam Generation 

      Buongiorno, J.; Todreas, N. E; Kazimi, Mujid S.; Czerwinski, K. R. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2001-03)
      The feasibility of a lead-bismuth (Pb-Bi) cooled fast reactor that eliminates the need for steam generators and coolant pumps was explored. The working steam is generated by direct contact vaporization of water and liquid ...

    • Plant Design and Cost Assessment of Forced Circulation Lead-Bismuth Cooled Reactor with Conventional Power Conversion Cycles 

      Dostal, Vaclav; Hejzlar, Pavel; Todreas, Neil E.; Kazimi, Mujid S. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2001-09)
      Cost of electricity is the key factor that determines competitiveness of a power plant. Thus the proper selection, design and optimization of the electric power generating cycle is of main importance. This report makes an ...

    • MCNP4B Modeling of Pebble-Bed Reactors 

      Lebenhaft, Julian Robert (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2001-10-15)
      The applicability of the Monte Carlo code MCNP4B to the neutronic modeling of pebble-bed reactors was investigated. A modeling methodology was developed based on an analysis of critical experiments carried out at the ...

    • A Super Critical Carbon Dioxide Cycle for Next Generation Nuclear Reactors 

      Dostal, Vaclav; Driscoll, Michael J.; Hejzlar, Pavel (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2004-03)
      A systematic, detailed major component and system design evaluation and multiple-parameter optimization under practical constraints has been performed of the family of supercritical CO[subscript 2] Brayton power cycles for ...

    • 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 ...

    • Analysis Of Flow Instabilities In Supercritical Watercooled Nuclear Reactors 

      Zhao, J.; Saha, P.; Kazimi, Mujid S. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2004-09)
      Near the supercritical thermodynamic point, coolant density is very sensitive to temperature which gives potential to several instabilities in the supercritical water-cooled nuclear reactors. The flow stability features ...

    • Stability Analysis of Supercritical Water Cooled Reactors 

      Zhao, J.; Saha, P.; Kazimi, Mujid S. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2005-09)
      The Supercritical Water-Cooled Reactor (SCWR) is a concept for an advanced reactor that will operate at high pressure (25MPa) and high temperature (500ºC average core exit). The high coolant temperature as it leaves the ...

    • Investigation of Pressure-Tube and Calandria-Tube Deformation Following a Single Channel Blockage Event in ACR-700 

      Gerardi, Craig Douglas; Buongiorno, Jacopo (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2005-11)
      The ACR-700 is an advanced pressure-tube (PT) reactor being developed by Atomic Energy of Canada Limited (AECL). As in conventional CANDU reactors, the PTs are horizontal. Each PT is surrounded by a calandria tube (CT), ...