The Reduced-Moderation Water Reactor (RMWR) is being developed at Japan Atomic Energy Agency and demonstration of the core heat removal performance is one of the most important issues. However, operation of the full-scale bundle experiment is difficult technically because the fuel rod bundle size is larger, which consumes huge electricity. Hence, it is expected to develop an analysis code for simulating RMWR core thermal-hydraulic performance with high accuracy. Subchannel analysis is the most powerful technique to resolve the problem. A subchannel analysis code NASCA (Nuclear-reactor Advanced Sub-Channel Analysis code) has been developed to improve capabilities of analyzing transient two-phase flow phenomena, boiling transition (BT) and post BT, and the NASCA code is applicable on the thermal-hydraulic analysis for the current BWR fuel. In the present study, the prediction accuracy of the NASCA code has been investigated using the reduced-scale rod bundle test data, and its applicability on the RMWR has been improved by optimizing the mechanistic constitutive models.

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