Richard R. Schultz

Validating and Verifying a New Thermal-Hydraulic Analysis Tool

The Idaho National Engineering and Environmental Laboratory (INEEL) has developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D{sup C}/ATHENA advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, three-dimensional analyses using Fluents CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D{sup C}/ATHENA. Both steady-state and transient calculations can be performed, using many working fluids and point to three-dimensional neutronics. A general description of the techniques used to couple the codes is given. The validation and verification (V and V) matrix is outlined. V and V is presently ongoing. (authors)

Preliminary Investigation of an Optimally Scramming Control Rod for Gas-Cooled Reactors

A passively safe control rod for gas-cooled reactors is proposed. This Optimally Scramming Control Rod (OSCR) is lifted out of the core region by the core coolant and descends back into the core when the coolant flow is not sufficient for core cooling purposes or in the event of depressurization. It is shown that for the current design of the OSCR, the reactor can be operated under normal lower power conditions down to about 80% of total power. It is also shown that cold shutdown can be achieved with rods of sufficiently low mass to allow naturally passive operation of the concept.Copyright

Saturated-Subcooled Stratified Flow in Horizontal Pipes

Advanced light water reactor systems are designed to use passive emergency core cooling systems with horizontal pipes that provide highly subcooled water from water storage tanks or passive heat exchangers to the reactor vessel core under accident conditions. Because passive systems are driven by density gradients, the horizontal pipes often do not flow full and thus have a free surface that is exposed to saturated steam and stratified flow is present.

Providing the Basis for Innovative Improvements in Advanced LWR Reactor Passive Safety Systems Design: An Educational R&D Project

This project characterizes typical two-phase stratified flow conditions in advanced water reactor horizontal pipe sections, following activation of passive cooling systems. It provides (1) a means to educate nuclear engineering students regarding the importance of two-phase stratified flow in passive cooling systems to the safety of advanced reactor systems and (2) describes the experimental apparatus and process to measure key parameters essential to consider when designing passive emergency core cooling flow paths that may encounter this flow regime. Based on data collected, the state of analysis capabilities can be determined regarding stratified flow in advanced reactor systems and the best paths forward can be identified to ensure that the nuclear industry can properly characterize two-phase stratified flow in passive emergency core cooling systems.

Stratified Flow Experiments Pertaining to Advanced LWR Reactor Passive Safety System Designs

During a loss-of-coolant-accident in advanced light water reactors, outside coolant enters the cold leg by gravity to cool the core. This coolant is at a substantially lower temperature and thus is heavier than the liquid in and from the reactor. Consequently, stratified flow may occur. A stratified flow may cause condensation-induced water hammer, and will influence the coolant flow behavior. Two sets of experiments are in progress to better understand stratified flow conditions that lead to water hammer, and the density stratification behavior. The first set uses air-oil-water as the test media. Its purposes are to conduct exploratory tests and to provide instruction an apparatus for education purposes. The second set of tests will use steam and water and, later, the refrigerant R123. This paper describes the exploratory test facility, gives a brief description of the facility that will be used for the steam-water and refrigerant tests, describes the overall test plan, and finally gives some preliminary results on the intrusion of a lighter liquid into a pipe against flow.Copyright