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

Now showing items 13-32 of 34

    • Gas Heat Transfer in a Heated Vertical Channel under Deteriorated Turbulent Heat Transfer Regime 

      Lee, Jeongik (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2007-04)
      Passive cooling via natural circulation of gas after a loss of coolant (LOCA) accident is one of the major goals of the Gas-cooled Fast Reactor (GFR). Due to its high surface heat flux and low coolant velocities under ...

    • General Analysis of Breed-and-Burn Reactors and Limited-Separations Fuel Cycles 

      Petroski, Robert C.; Forget, Benoit (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2011-02)
      A new theoretical framework is introduced, the “neutron excess” concept, which is useful for analyzing breed-and-burn (B&B) reactors and their fuel cycles. Based on this concept, a set of methods has been developed which ...

    • 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 ...

    • 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), ...

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

      Gerardi, C.; Buongiorno, J. (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), ...

    • 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 ...

    • Methods for Comparative Assessment of Active and Passive Safety Systems 

      Oh, Jiyong; Golay, Michael W. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2008-02)
      Passive cooling systems sometimes use natural circulation, and they are not dependent on offsite or emergency AC power, which can simplify designs through the reduction of emergency power supplying infrastructure. The ...

    • 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 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 ...

    • 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 ...

    • 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 ...

    • 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. ...

    • 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 ...

    • 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 ...

    • Stability Analysis of Natural Circulation in BWRs at High Pressure Conditions 

      Hu, Rui; Kazimi, Mujid S. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2007-10)
      At rated conditions, a natural circulation boiling water reactor (NCBWR) depends completely on buoyancy to remove heat from the reactor core. This raises the issue of potential unstable flow oscillations. The objective ...

    • 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 ...

    • 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 ...

    • Thermal-Hydraulic Analysis of Innovative Fuel Configurations for the Sodium Fast Reactor 

      Memmott, Matthew J.; Hejzlar, Pavel; Buongiorno, Jacopo (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2009-08)
      The sodium fast reactor (SFR) is currently being reconsidered as an instrument for actinide management throughout the world, thanks in part to international programs such as the Generation-IV and especially the Global ...

    • 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 ...

    • Use of Response Surface for Evaluation of Functional Failure of Passive Safety System 

      Fei, Tingzhou; Golay, Michael W. (Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program, 2010-03)
      Passive safety systems are more vulnerable to their environment and initial condition due to the typical low driving forces, e.g., that of the natural circulation. We investigate the merits of different methods for ...