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題 名 | The Survey of Critical Heat Flux Correlations=臨界熱傳量經驗公式之檢視 |
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作 者 | 吳邦彥; | 書刊名 | 德霖學報. 理工類 |
卷 期 | 16 2002.06[民91.06] |
頁 次 | 頁42-66 |
分類號 | 449.2 |
關鍵詞 | 臨界熱傳量; 乾化; 燒毀; 偏離核沸騰; 沸騰轉捩點; CHF; Critical heat flux; Dryout; Burnout; DNB; Boiling crisis; |
語 文 | 英文(English) |
中文摘要 | 水冷式核子反應器的燃料鞘溫度經常接近水的飽和溫度,當電力無預警增加或者是冷卻水流流量或壓力減小時,將使加熱表面不再持續的與冷卻水接觸,此時會發生熱傳劣化現象。此熱傳劣化現象一般稱為沸騰轉捩點、或乾化、或燒毀、或偏離核沸騰,此時相對應的熱傳量稱為臨界熱傳量,簡稱CHF。臨界熱傳量即為發生乾化現象時的熱傳量。當二相管狀流液態薄膜不再覆蓋在散熱面上時,就產生乾化現象。此現象將造成散熱表面溫度微小但快速的增加,同時將造成乾燥點產生與消滅。因此,當熱量超過臨界熱傳量時,將造成散熱面的失效。水冷式核子反應器的散熱面的失效,可能造成中心燃料棒的熔解。為避免造成產生過高的燃料鞘溫度,一般有兩種方法:(1)保持散熱量低於臨界熱傳量,或(2)於特定的操作情況,使後乾化(post-dryout)熱傳能有效運作,以致於燃料鞘溫度保持在可接受的溫度範圍內。要決定一個反應器的熱傳極限,必須經由實驗方法求得臨界熱傳量。基於不均勻的熱量分布與真實反應器複雜的幾何形狀,現階段欲正確量測求得臨界熱傳量是有困難的。本研究搜尋許多的論文與技術報告,整理出許多預測求取臨界熱傳量的經驗公式。縱使於實際的使用上沒有一個公式適用於所有的情況,但是部分的公式可以作為設計與操作時的參考。 |
英文摘要 | Fuel sheath temperatures in water-cooled nuclear reactors are usually near the saturation temperature of water. During an accidental increase in power or decrease in flow and pressure, heat transfer deterioration can occur so that the heated surface temperature can no longer support continuous liquid contact. This phenomenon is usually referred to as the boiling crisis or dryout, burnout and departure from nucleate boiling (DNB) and the corresponding heat flux as the critical heat flux or CHF. CHF is that heat at which dryout occurs. Dryout occurs when the liquid film covering the heat surface in two-phase annular flow breaks down. It causes small but rapid rises in surface temperature corresponding to the appearance and disappearance of the dry patches. Failure of the heated surface may occur once the CHF is exceeded. In the case of water-cooled reactors once fuel element surface heat flux exceeds the CHF the fuel element may experience fuel centerline melting. High sheath temperatures can be avoided by either operating at heat flux levels below the critical heat flux or by operating at conditions where the post-dryout heat transfer is reasonably effective in keeping the fuel sheath temperature at moderate level. To determine the heat transfer limits of a reactor channel, the CHF must be determined experimentally. Because of the non-uniform flux distribution and the complicated geometry of real reactor fuel it is impossible to accurately predict CHF at this stage. In this report, many correlations are presented by surveying the published papers and technical reports. Although none of them is suitable for every case in reality, some of them could be the design and operating references. |
本系統中英文摘要資訊取自各篇刊載內容。