[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 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] 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] Following postulated loss-of-decay-heat removal accidents in a boiling water reactor (BWR), continued steam generation in the core could lead to containment overpressurization. The Emergency Procedure Guidelines (EPG), developed jointly by the BWR Owners Group and General Electric Company, recommend operator action to vent the primary containment, as the last resort, to avoid adverse consequence to the containment or potentially to the core. The EPGs provide recommendations to be used by individual utilities in development of their plant's unique emergency operating procedures (EOPs). In formulation of the EOPs for containment venting, a number of factors must be assessed to assure that the operator actions are taken at the proper time and that the minimum consequence of venting results. Among these factors, induced pool-swell loads on containment were postulated due to such events as downcomer clearing and bulk nucleation. This paper presents the results of a study, using computer code TRACB04 of pool swelling induced by venting the containment at high pressure. TRACB04, an enhanced version of TRAC, is a best-estimate computer code for the analysis of BWR transients

[en] TRACB02 is a computer code developed at General Electric Company (GE) in a collaboration with the Idaho National Engineering Laboratory for best-estimate analysis of the thermal-hydraulic conditions in a boiling water reactor (BWR). This development effort was jointly funded by the US Nuclear Regulatory Commission/Electric Power Research Institute/GE. TRAC features a high degree of modularity to allow for the simulation of almost any system. The basic physical phenomena are described by the conservation equations for mass, momentum, and energy. The heat transfer mechanisms modeled include single-phase convection, nucleate and sub-cooled boiling, boiling transition and rewet, post-dryout heat transfer, thermal radiation, and interfacial heat transfer to droplets, bubbles, and films. The heat transfer models are tied to the interfacial shear model through a consistent flow regime map. The assessment was performed by simulating a set of separate effects tests with TRACB02 and then comparing the results with data. Results of the assessment studies are summarized

[en] This paper examines the influence of spacers on the critical power of a BWR bundle. COBRAG, a detailed subchannel analysis code developed at GE Nuclear Energy, is used for the study. The study includes the effects of the spacer design (grid-type vs tube-type), physical dimensions (material thickness and height) and location (spacer pitch). Results from the study are compared to the critical power measurements from the ATLAS Test Facility at GE Nuclear Energy. The critical power trends for these parameters are also presented. Generally, good agreement between data and COBRAG predictions has been obtained, and demonstrates the capability of COBRAG to capture the effects of spacers on the critical power

[en] Predictions of critical power by COBRAG based on a two-fluid, multi-field model were compared against the data collected at the ATLAS test facility at GE Nuclear Energy. Results of the comparisons are good with a relative percentage error generally less than 5%. The predicted trends in critical power versus some important physical parameters arc also found to be in close agreement with experiments. (author)

[en] The ability to accurately predict the time required to reach a boiling transition in postulated BWR (boiling water reactor) loss-of-coolant accidents is an important step in analyzing the system response to these transients. As part of the effort to develop an advanced BWR system analysis computer code capable of simulating these postulated transients, a study has been undertaken to improve the boiling transition calculations associated with LOCA-type (loss-of-coolant accidents) phenomena. 16 refs