Engineering topics
Publications and source records attributed to Anderson, Mark H.
Update on the use of fiber optic temperature sensing technologies for examining thermal stratification in a liquid sodium facility. Research discusses techniques for measuring temperature with optical fibers as well as presenting data from research.
Sodium fast reactors are poised to be a leading candidate for the next generation of commercial reactor deployment. Before licensing can be completed it is important that information be gathered which examines reactor transients during loss of flow accidents. During loss of flow accidents thermally stratified layers can form in the reactor pool. The formation of thermal stratification in the primary pool of a sodium fast reactor during a loss of flow accident can cause cyclic thermal fatigue on the reactor vessel potentially leading to crack formation and weld failure. Computational models have been built to analyze the formation and persistence of thermal stratification during accident transients, but are not currently validated by experimental data. A thermal stratification research facility has been developed and tested to research thermal stratification with loss of flow type conditions. Data from the experiments will be used to validate computational models. The facility can simulate both protected and unprotected loss of flow accident transients in the temperature range of 200-325 [C] with flow rates from 2 gpm to 20 gpm.
Commercial utilization of the diffusion-bonded heat exchanger (DBHE) for high-temperature applications has prompted an ASME Boiler and Pressure Vessel Code Case development effort to employ these components in Section III (nuclear plant) applications. As part of this effort, a series of DBHE have been fabricated. These heat exchangers will be among the first printed-circuit geometries to be tested for functionality in a molten salt environment. These 15-25kW test specimens will be evaluated under steady-state conditions to compare pressure drop and terminal temperatures measurements with those predicted by existing flow and heat transfer correlations. The test units will then be subjected to thermal transients similar to those expected in Generation IV reactor systems by varying the hot and cold fluid mass flow rates and by changing inlet temperatures. Data collected from these tests will narrow the technical gaps related to two-fluid PCHE and identify special design requirements of compact heat exchangers containing primary reactor coolant. Descriptions of one of two test units and the testing plan are discussed here.