第四代核反應器是新型態的核反應器，可望提供永續、穩定且經濟的能源，並大幅度改善電廠安全。超臨界水反應器是其中之一，具有高熱功率，系統簡單等特質，加上超臨界水燃煤電廠、壓水式核反應器與沸水式核反應器的豐富運轉資訊作為參考，為具有發展潛力的反應器。
本研究的目標，是使用RELAP5/MOD3程式，建立超臨界水反應器之穩態輸入數據以及暫態分析。RELAP5/MOD3是愛達荷國家工程實驗室(INEL)與美國核能管制委員會(NRC)合作發展的熱水流分析程式，為輕水式核反應器所用。本論文之穩態輸入參數，根據美國愛達荷國家實驗室於2005年一月發表的報告所提供之超臨界水反應器設計參數。模擬範圍包含反應器壓力槽及內部組件、熱端管路、冷端管路部份。
本論文暫態分析選擇喪失主要飼水事故(Loss of Main Feedwater, LOMF)。在適當的安全系統設計之下，暫態發生後的燃料護套最大溫度可維持在法規限值840℃之下，然而分析程式RELAP5/MOD3的熱傳模式有缺陷，導致無法預測後續暫態反應。

The Supercritical Water Reactor (SCWR), one of the six possible types of Gen-IV reactor, is a Light Water Reactor (LWR) operating at supercritical pressure. It has the advantage of high thermal efficiency, plant system simplifications and proven technologies.
In this study, RELAP5/MOD3 code is used to simulate the steady state of the SCWR and to analyze the transients. The components considered in the input deck include SCWR core, 2 loop hot legs, cold legs, and safety components.
Analyses have shown that a transient caused by a Loss of Main Feedwater (LOMF) event can be a challenging event for a SCWR. The propose of the safety system is to make the peak cladding temperature remain less than the transient limit 840℃. Due to some defect of the heat transfer model of RELAP5/MOD3 code, the transient analyses are not complete.

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