[en] CORECOOL, Convection and Radiation Emergency Cooling, is a model for the two-phase flow and heat transfer in a fuel element during the core heat-up phase following a loss of coolant accident. The model for the two-phase flow is based on a solution of the conservation equations with a separate description of the two phases and thermodynamic non-equilibrium. The flow-regimes considered are drop flow and film flow. The heat transfer consists of convection, sputtering and radiation heat transfer. The documentation of CORECOOL consists of four parts: I) model description, II) programme description (COMMERCIAL), III) users guide (COMMERCIAL) IV) verification (COMMERCIAL). CORECOOL is a joint project between Risoe National Laboratory Denmark and General Electric Company, San Jose, USA. (author)

[en] A complete technical basis for implementation of the 3-D fast numerics in TRACB04 is presented. The 3-D fast numerics is a generalization of the predictor/corrector method previously developed for the 1-D components in TRACB. 20 figs

[en] A process is described for providing a radiation heat sink for fuel bundles having a large water moderator tube in the event of a loss of coolant accident the fuel bundles having an upper tie plate, a lower tie plate, a channel surrounding and connecting the tie plate, a plurality of fuel rods supported between the tie plates and within the channels in side by side upstanding relation; a large water moderator tube having at least twice the diameter of the fuel rods. The process consists of: spraying core cooling spray in an evenly divided flow over the upper tie plate; collecting core cooling spray at an uper end of the large water moderator tube; and distributing the core cooling spray circumferentially along the inner surfaces of the large water moderator tube in a downward flow separating the flow of the core cooling spray from the flow of steam resulting from the flashing of water to steam within the moderator tube

[en] TRAC (Transient Reactor Analysis Code) is a computer code for best estimate analysis of the thermal hydraulic conditions in a reactor system. The constitutive correlations for shear and heat transfer in the boiling water reactor (BWR) version of TRAC are described. A new model, that accounts for the effect of phase and velocity profiles, has been developed for the interfacial shear and a new set of constitutive correlations are derived. Improvements have been made to the heat transfer in the area of subcooled boiling, boiling transition, and thermal radiation

[en] TRAC (Transient Reactor Analysis Code) is a computer code for best estimate analysis of the thermal hydraulic conditions in a reactor system. The constitutive correlations for shear and heat transfer developed for the Boiling Water Reactor (BWR) version of TRAC are described. A universal flow regime map has been developed to tie the regimes for shear and heat transfer into a consistent package. New models in the areas of interfacial shear, interfacial heat transfer and thermal radiation have been introduced. Improvements have also been made to the constitutive correlations and the numerical methods. All the models have been implemented into the GE version TRACB02 and extensively tested against data

[en] The phenomenon of film boiling heat transfer for high wall temperatures has been investigated. Based on the assumption of laminar flow for the film, the continuity, momentum, and energy equations for the vapor film are solved and a Bromley-type analytical expression for the heat transfer coefficient versus the film length is obtained. The Helmholtz instability for the steam-liquid interface is analyzed, and it is shown that films beyond a certain length are unstable. Assuming the most unstable wavelength for disturbances at the steam-liquid interface is a reasonable expression for the film length, an average film boiling heat transfer coefficient is obtained

[en] The TRACB04 computer code has been developed under the model development tasks in the FIST Program. This report describes two developmental assessment calculations performed on BWR plants with TRACB04. A BWR/2 Design Basis Accident (DBA) including the containment response and a BWR/4 DBA with Low Pressure Coolant Injection (LPCI) water injected into the lower plenum were calculated and results of these cases were documented. These cases serve to test some of the new features of the TRACB04 (air field, containment model, ''water packing'' fixes and faster numerics in the three dimensional vessel component) and to demonstrate that the code has been assembled properly. They also provide best estimate LOCA results for the two plant types

[en] The Advanced Boiling Water Reactor will be the next generation BWR in Japan. This paper discusses the enhanced features of the ABWR and compares the ABWR ECCS network to previous BWR designs. The limiting LOCA case for the ABWR which was analyzed with the recently approved SAFER LOCA evaluation model, is described in detail. The results show that no fuel uncovery should occur for any possible break and assumed single failure. These results are also compared to the limiting LOCA analysis for previous BWR designs. To help qualify the SAFER model, the SAFER results are compared to corresponding TRAC results for a special case. This comparison demonstrates that the simpler SAFER model adequately predicts the major phenomena which occur during the accident. Therefore, it is concluded that SAFER can be used to perform ABWR LOCA analyses for both design basis events and nonstandard sensitivity studies

[en] TRAC-BWR (Transient Reactor Analysis Code) is a computer code for best estimate analysis of the thermal hydraulic conditions in a Boiling Water Reactor system. In this report, the development of new models and the implementation of the balance of plant models leading to the creation of the TRACB04 version of the code, is described. The new models include an improved model for boron transport which accounts for non-uniform mixing and stratification, and a model for the interfacial heat transfer at two-phase levels. The balance of plant models (turbine, containment and heat exchanger) developed at INEL were evaluated, adapted, and implemented into TRACB04 to provide complete transient analysis capability. In addition, a model for air or a noncondensible gas as an additional field in the system of equations was adapted to the two step numerical method and incorporated into TRACB04

[en] Considerable experimental and analytical work have been done to evaluate the heat transfer phenomena in BWR fuel elements during spray cooling conditions. Two types of experimental studies have been performed. Basic separate-effect experiments have given some detailed information on the individual phenomena during BWR spray cooling conditions. Experiments with full scale BWR fuel element simulators have given extensive information on the integrated performance of the fuel element under representative spray cooling conditions. The experiments have been complemented with both semiempirical models and detailed mechanistic models based on physical descriptions of the individual two-phase flow and heat transfer phenomena in the bundle. These methods are described and it is concluded that a comprehensive amount of information and phenomena understanding on BWR spray cooling heat transfer exists, and that it is presently possible to perform accurate and detailed calculations of core heat-up transients for a BWR-LOCA. 68 references