Dispersed-flow film boiling data were obtained in a large rod bundle (8 × 8) under steady-state and transient conditions with upward flowing high-pressure, high-temperature water. The bundle is equipped with detailed thermometry, and has geometry typical of later generation pressurized water reactors with 17 × 17 fuel assemblies. Comparisons with the data to empirical correlations commonly used to predict heat transfer in dispersed flow indicate that the Dougall-Rohsenow and Groeneveld-Delorme correlations overpredict and underpredict heat transfer, respectively, while the Groeneveld 5.7 and Condie-Bengston IV correlations perform reasonably well.Spacer grids are shown to cause rod surface temperature depressions of up to 100 K from the upstream to downstream side of the grid. Grid effects persist for 20 to 30 hydraulic diameters downstream of the grid.

A series of steady-state film boiling experiments have been conducted to show the effect of spacer grids on rod bundle heat transfer. Experiments were performed at the Oak Ridge National Laboratory in the Thermal-Hydraulic Test Facility, a pressurized water loop containing a 64-rod bundle, of which 60 rods are electrically heated. The bundle is equipped with detailed thermometry around two grids, which allows grid heat transfer effects to be studied. Rod surface temperature data show a 75 to 150 K temperature difference between measurements upstream and downstream of the grids, while heat transfer coefficients increase across the grids by 20%. Twenty to thirty hydraulic diameters are required for these effects to dissipate.

Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) Test 3.03.6AR. This test was conducted by members of the ORNL Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on May 21, 1980. Objective was to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small and large break loss-of-coolant accidents. Test 3.03.6AR was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.03.6AR available. Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.

An advanced methodology for performing systematic uncertainty analysis of time-dependent nonlinear systems is presented. This methodology includes a capability for reducing uncertainties in system parameters and responses by using Bayesian inference techniques to consistently combine prior knowledge with additional experimental information. The determination of best estimates for the system parameters, for the responses, and for their respective covariances is treated as a time-dependent constrained minimization problem. Three alternative formalisms for solving this problem are developed. The two "off-line" formalisms, with and without "foresight" characteristics, require the generation of a complete sensitivity data base prior to performing the uncertainty analysis. The "online" formalism, in which uncertainty analysis is performed interactively with the system analysis code, is best suited for treatment of large-scale highly nonlinear time-dependent problems.This methodology is applied to the uncertainty analysis of a transient upflow of a high pressure water heat transfer experiment. For comparison, an uncertainty analysis using sensitivities computed by standard response surface techniques is also performed. The results of the analysis indicate the following. Major reduction of the discrepancies in the calculation/experiment ratios is achieved by using the new methodology.Incorporation of in-bundle measurements in the uncertainty analysis significantly reduces system uncertainties.Accuracy of sensitivities generated by response-surface techniques should be carefully assessed prior to using them as a basis for uncertainty analyses of transient reactor safety problems.Conclusions about the future applicability of the uncertainty analysis methodology presented in this work are also discussed.

Six additional bundle uncovery/recovery tests were performed in the Thermal-Hydraulic Test Facility during January, bringing the total number of these tests to eight. Data taken during the tests were found to be contaminated by numerous spurious spikes. Work to remove the spurious spikes is under way. Posttest analysis of the tests is approx.20% completed. The recovery portion of one of the tests will be analyzed by COBRA/TRAC, currently being developed by Pacific Northwest Laboratories (PNL). Work to debug the code for this application is in progress at PNL. The uncovery/recovery tests apparently caused damage to the 0-rings that form part of the loop pressure boundary. Refurbishment of the 0-ring seal system is being performed concurrently with scheduled loop modifications that include installation of ten in-bundle differential pressure instruments. Design, procurement, and fabrication of the in-bundle gamma densitometer system are continuing on schedule.

Rod bundle heat transfer data near spacer grids has been investigated in the Thermal Hydraulic Test Facility (THTF). Steady state, upflow film boiling tests show the grid's effect on both local surface temperatures and heat transfer coefficients.

The input data used with the computer program RELAP4/MOD5 to model the Thermal Hydraulic Test Facility (THTF), as part of the ORNL PWR-BDHT Program analysis effort, are described. A model of the whole THTF system and a three-channel component model of the THTF test section are covered. Emphasis is placed on the original data sources and analysis techniques used in developing input model data.

Heat transfer phenomena are analyzed for test series 100 of the Thermal-Hydraulic Test Facility, part of the Pressurized-Water Reactor Blowdown Heat Transfer Separate-Effects Program. Heater rod surface temperatures are found to be sensitive to relatively small variations in flow. The mechanisms causing departure from nucleate boiling and rewetting are analyzed. Comparisons are made between heater rod surface temperatures calculated from thermocouple responses and surface temperatures produced by the thermal-hydraulic transient simulator RELAP4/MOD5 (update 2). The code's predictions are relatively accurate in the lower region of the test section and considerably high in the upper region. The procedures in the code which produce these results are discussed in detail.