SCALING OF COMPLEX PHENOMENA IN SYSTEM THERMAL-HYDRAULICS

A methodology is described that can be used for the extrapolation of t hermal-hydraulic phenomena measured in differently scaled integral tes t facilities to nuclear reactor plant conditions. The use of a system code in this context is confirmed to be of fundamental importance, pro vided that the code's scaling capability has been demonstrated. The st arting data base for the proposed study consists of the measured quant ities and corresponding RELAP5/MOD2 code calculation results related t o a boiling water reactor small-break loss-of-coolant accident (SBLOCA ) counterpart test activity, a pressurized water reactor (PWR) natural -circulation type test activity, and a PWR SBLOCA counterpart test act ivity. The proof that this methodology can be used for evaluating unce rtainties in predicting transient behavior in nuclear power plants is the main result of this study. Data have been obtained that give a val ue of the foreseeable error ranges in the provision of plant behavior in the three cases considered.