Browsing Advanced Nuclear Power Technology Program (ANP) - Technical Reports by Title
Now showing items 9-28 of 34
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Design of a Functionally Graded Composite for Service in High Temperature Lead and Lead-Bismuth Cooled Nuclear Reactors
(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 Compact Intermediate Heat Exchangers for Gas Cooled Fast Reactors
(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 ... -
Effective Thermal Conductivity of Prismatic MHTGR Fuel
(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
(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 ... -
Gas Heat Transfer in a Heated Vertical Channel under Deteriorated Turbulent Heat Transfer Regime
(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
(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
(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
(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
(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
(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
(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
(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
(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
(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
(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
(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
(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
(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
(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
(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 ...