【摘要】：鉛基堆具有良好的中子學(xué)性能、熱工水力學(xué)性能、化學(xué)性能和固有安全特性而成為極具發(fā)展?jié)摿Φ牡谒拇四芟到y(tǒng)。在鉛基堆停堆后,堆芯活性區(qū)產(chǎn)生的大量余熱嚴(yán)重威脅著反應(yīng)堆的安全。鉛基堆非能動(dòng)余熱排出系統(tǒng)(PRHRS)的設(shè)計(jì)研究,對于提高鉛基堆在事故工況下的同有安全性,具有重要理論意義和應(yīng)用價(jià)值。本文以1000 MWth鉛基堆為研究對象,從鉛鉍換熱模型開發(fā)與驗(yàn)證、三種PRHRS的設(shè)計(jì)與性能分析、參數(shù)敏感性分析和性能對比分析四個(gè)方面開展了鉛基堆PRHRS的設(shè)計(jì)研究。相關(guān)研究成果可為鉛基堆非能動(dòng)余熱排出系統(tǒng)的設(shè)計(jì)提供了經(jīng)驗(yàn)和參考,對我國自主掌握鉛基堆非能動(dòng)余熱排出的關(guān)鍵技術(shù)有所借鑒。主要研究成果如下:1)傳統(tǒng)壓水堆的換熱模型不適用于鉛鉍換熱計(jì)算,本文基于OECD/NEA LACANES計(jì)劃中有關(guān)鉛鉍換熱的驗(yàn)證工作,在熱工水力學(xué)分析程序RELAP5/MOD4.0基礎(chǔ)上開發(fā)了鉛鉍換熱模型,并通過單管測試驗(yàn)證了此鉛鉍換熱模型開發(fā)的正確性,建立了鉛基堆熱工水力學(xué)設(shè)計(jì)與分析的工具基礎(chǔ)。2)根據(jù)非能動(dòng)安全設(shè)計(jì)理念,分別設(shè)計(jì)了基于蒸汽發(fā)生器(SG)、獨(dú)立熱交換器(IHEX)和反應(yīng)堆容器(RV)的三種PRHRS,并使用RELAP5/MOD 4.0程序分析了全廠斷電事故下三種PRHRS的余熱排出性能。研究結(jié)果表明:基于SG、IHEX和RV的PRHRS余熱排出速率分別在事故后0.6 h、5 h和48 h超過堆芯衰變余熱,包殼峰值溫度分別達(dá)到490 ℃、562 ℃和743 ℃,低于包殼溫度限值(750 ℃),鉛基堆處于安全狀態(tài)。三種PRHRS均可應(yīng)對全廠斷電這類超設(shè)計(jì)基準(zhǔn)事故,滿足設(shè)計(jì)目標(biāo)。3)對三種PRHRS參數(shù)敏感性進(jìn)行了分析,研究結(jié)果表明:基于SG和IHEX的PRHRS,通過增大換熱面積和減小系統(tǒng)響應(yīng)時(shí)間可以增強(qiáng)PRHRS的余熱排出性能;而增加冷熱芯高度差和減小回路阻力對PRHRS的余熱排出性能影響較小�；赗V的PRHRS,通過增大煙囪高度和換熱面積、減小間隙寬度可以增強(qiáng)PRHRS的余熱排出性能;而增加壁面發(fā)射率和減小系統(tǒng)響應(yīng)時(shí)間對PRHRS的余熱排出性能影響較小。4)針對三種PRHRS的性能進(jìn)行對比分析,結(jié)果表明:基于SG的PRHRS排熱能力最強(qiáng),更適合用于大功率鉛基堆/商業(yè)堆,或在事故前期投入運(yùn)行,帶走停堆初期的大量余熱;基于IHEX的PRHRS排熱能力較強(qiáng),更適合用于中等功率鉛基堆/示范堆,或在事故中期投入運(yùn)行;而基于RV的PRHRS排熱能力較弱,更適合用于小功率鉛基堆/研究堆,或作為事故后期長期排熱的余熱排出系統(tǒng)。同時(shí),在實(shí)際鉛基堆PRHRS設(shè)計(jì)中,為提高余熱排出的可靠性,建議采取兩種以上的非能動(dòng)余熱排出系統(tǒng)。
[Abstract]:Lead based reactor is the fourth generation nuclear power system with great potential for development due to its good neutronic properties, thermal hydraulic properties, chemical properties and inherent safety properties. After the lead-based reactor is shut down, the reactor safety is seriously threatened by a large amount of residual heat generated in the reactor core reactive zone. The design and study of (PRHRS) for lead based reactor passive residual heat removal system is of great theoretical significance and application value for improving the safety of lead based reactor under accident conditions. In this paper, 1000 MWth lead based reactor is taken as the research object. The design and verification of lead bismuth heat transfer model, the design and performance analysis of three kinds of PRHRS, the parameter sensitivity analysis and the performance contrast analysis are carried out. The design and research of lead based reactor PRHRS are carried out. The related research results can provide experience and reference for the design of passive residual heat removal system for lead based reactor, and can be used for reference to master the key technology of passive residual heat removal of lead based reactor in China. The main research results are as follows: 1) the traditional heat transfer model of PWR is not suitable for the calculation of lead and bismuth heat transfer. This paper is based on the verification of lead and bismuth heat transfer in OECD/NEA LACANES plan. A lead bismuth heat transfer model was developed on the basis of thermal hydraulic analysis program RELAP5/MOD4.0, and the correctness of the model was verified by a single tube test. The tool foundation of thermal hydraulics design and analysis for lead based reactor is established. 2) according to the concept of inactive safety design, three kinds of PRHRS, based on (SG), independent heat exchanger (IHEX) and reactor container (RV) are designed, respectively. The residual heat removal performance of three kinds of PRHRS under power failure accident was analyzed by RELAP5/MOD 4.0 program. The results show that the residual heat removal rate of PRHRS based on SG,IHEX and RV exceeds the decay residual heat of core at 0.6 h and 48 h, respectively, and the peak temperature of cladding reaches 490 鈩，

本文編號：2366099