High Temperature Stress Analysis on 61-pin Test Assembly for Reactor Core Sub-channel Flow Test

In this study, a high temperature heat transfer and stress analysis of a 61-pin test fuel assembly scaled down from the full scale 217-pin sub-assembly was conducted. The reactor core subchannel flow characteristic test will be conducted to evaluate uncertainties in computer codes used for reactor core thermal hydraulic design. Stress analysis for a 61-pin fuel assembly scaled down from Prototype Generation IV Sodium-cooled Fast Reactor was conducted and structural integrity in terms of load controlled stress limits was conducted. In this study, The evaluations on load-controlled stress limits for a 61-pin test fuel assembly to be used for reactor core subchannel flow distribution tests were conducted assuming that the test assembly is installed in a Prototype Generation IV Sodium-cooled fast reactor core. The 61-pin test assembly has the geometric similarity on P/D and H/D with PGSFR and material of fuel assembly is austenitic stainless steel 316L. The stress analysis results showed that 4.05MPa under primary load occurred at mid part of the test assembly and it was shown that the value of 4.05Mpa was far smaller than the code allowable of 127MPa. , it was shown that the stress intensity due to due to primary load is very small. The stress analysis results under primary and secondary loads showed that maximum stress intensity of 84.08MPa occurred at upper flange tangent to outer casing and the value was well within the code allowable of 268.8MPa. Integrity evaluations based on strain limits and creep-fatigue damage are underway according to the elevated design codes