DOE OSTI2010
Ex-vessel debris coolability is an important light water reactor (LWR) safety issue. For existing plants, resolution of this issue will confirm the technical basis for severe accident management guidelines (SAMGs). For new reactors, understanding this issue will help confirm the effectiveness of the design and implementation of new accident mitigation features and severe accident management design alternatives (SAMDAs). The first OECD-MCCI program conducted reactor material experiments focused on achieving the following technical objectives: i) provide confirmatory evidence for various cooling mechanisms through separate effect tests and data for severe accident code model development, and ii) provide longer-term two-dimensional core-concrete interaction data for code assessment and improvement. As one of the steps in satisfying these program objectives, the Management Board (MB) approved the conduct of a fourth longer-term 2-D Core-Concrete Interaction (CCI) experiment designed to provide information in several areas, including: i) lateral vs. axial power split during dry core-concrete interaction, ii) integral debris coolability data following late phase flooding, and iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. The first three tests investigated the interaction of fully oxidized PWR core melt compositions with specially designed concrete test sections that were initially 50 cm x 50 cm in cross-sectional area. Both siliceous and limestone/common sand concrete types were addressed in the test matrix. The fourth test that is addressed in this report, CCI-4, utilized a similar test apparatus and experiment boundary conditions, including a limestone/common sand concrete test section. However, instead of a fully oxidized PWR melt, CCI-4 utilized a partially oxidized BWR core melt containing ~ 8 wt % structural steel constituents. Thus, this test evaluated the effect of elevated melt metal content on two-dimensional cavity erosion and debris cooling behavior. The experiment was conducted on May 26, 2007, and early experiment results were reported in a quick-look data report. The objective of this final report is to document both the thermalhydraulic and posttest debris examination results for this test. To this end, a summary description of the test apparatus is provided first, followed by descriptions of the test operating procedure and experimental results that include both thermalhydraulic data and the findings from the posttest examinations. Overall specifications for CCI-4 are provided.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗