系統(tǒng)安全分析程序在超臨界水冷堆和鈉冷快堆上的適用性研發(fā)與應(yīng)用
發(fā)布時(shí)間：2018-09-12 07:14
【摘要】：系統(tǒng)安全分析程序（以下簡(jiǎn)稱(chēng)系統(tǒng)程序）是對(duì)反應(yīng)堆系統(tǒng)進(jìn)行瞬態(tài)和事故安全分析的重要工具。在近十年來(lái)第四代先進(jìn)核能系統(tǒng)研發(fā)熱潮背景下，國(guó)際核能界針對(duì)第四代反應(yīng)堆系統(tǒng)程序開(kāi)展了大量研究工作。由于輕水堆系統(tǒng)程序發(fā)展成熟，經(jīng)過(guò)了全面系統(tǒng)的驗(yàn)證和廣泛的應(yīng)用，其中一些被加以修改以適用于第四代核能系統(tǒng)。這些修改后的程序的適用性及可靠性還有待進(jìn)一步研究。 本文針對(duì)第四代反應(yīng)堆中的超臨界水冷堆和鈉冷快堆進(jìn)行系統(tǒng)程序的模型開(kāi)發(fā)、驗(yàn)證與應(yīng)用研究。首先對(duì)輕水堆系統(tǒng)程序ATHLET在這兩種第四代堆型上的適用性進(jìn)行評(píng)價(jià)，在此基礎(chǔ)上對(duì)ATHLET程序進(jìn)行二次模型開(kāi)發(fā)，即開(kāi)發(fā)適用于超臨界水和鈉兩種流體的相關(guān)計(jì)算模型，獲得多流體系統(tǒng)程序ATHLET-MF（Multi-Fluid）；其次，對(duì)修改后的ATHLET-MF程序進(jìn)行初步驗(yàn)證，并采用法國(guó)PHENIX鈉冷快堆的自然對(duì)流實(shí)驗(yàn)對(duì)程序進(jìn)行評(píng)估；最后，將ATHLET-MF程序應(yīng)用于超臨界水冷堆燃料性能驗(yàn)證實(shí)驗(yàn)回路的安全分析，并提出對(duì)該實(shí)驗(yàn)回路的安全系統(tǒng)設(shè)計(jì)的改進(jìn)意見(jiàn)。 論文的主要工作包括： 1. ATHLET程序的適用性評(píng)價(jià)與多流體模型開(kāi)發(fā)：（1）就ATHLET程序?qū)ΤR界水冷堆和鈉冷快堆的適用性進(jìn)行分析；（2）增加超臨界水的物性計(jì)算、傳熱、壓降、臨界流模型；（3）增加鈉的物性計(jì)算、傳熱、壓降模型；（4）引入多流體通用接口，將針對(duì)超臨界水和鈉兩種流體的程序擴(kuò)展整合為多流體模型，獲得ATHLET-MF程序版本。 2. ATHLET-MF程序多流體模型的初步驗(yàn)證與評(píng)估：（1）超臨界水冷堆擴(kuò)展功能的初步驗(yàn)證，包括與理論計(jì)算對(duì)比、程序比對(duì)驗(yàn)證；（2）鈉冷快堆擴(kuò)展功能的評(píng)估，即PHENIX反應(yīng)堆自然對(duì)流實(shí)驗(yàn)驗(yàn)證。結(jié)果表明：ATHLET-MF程序?qū)ΤR界水冷堆和鈉冷快堆系統(tǒng)的瞬態(tài)模擬具有良好的適用性和可靠性。 3. ATHLET-MF程序多流體模型的應(yīng)用，即超臨界水冷堆燃料性能驗(yàn)證實(shí)驗(yàn)回路的安全分析：（1）計(jì)算分析了5類(lèi)設(shè)計(jì)基準(zhǔn)事故，包括冷卻劑喪失、主泵卡軸、實(shí)驗(yàn)回路斷電、冷卻劑流道堵塞、以及由于壓力管內(nèi)部構(gòu)件破裂導(dǎo)致的冷卻劑旁流事故；（2）針對(duì)初始安全系統(tǒng)設(shè)計(jì)的缺陷提出改進(jìn)措施：增加了2個(gè)安全信號(hào)，提出長(zhǎng)期余熱排出方案并進(jìn)行了可行性分析；（3）對(duì)安全系統(tǒng)的部分設(shè)計(jì)參數(shù)進(jìn)行了敏感性分析，提出對(duì)安全系統(tǒng)控制信號(hào)及設(shè)備參數(shù)的建議。 本文旨在開(kāi)發(fā)和驗(yàn)證適用于超臨界水冷堆和鈉冷快堆的系統(tǒng)安全分析工具，為超臨界水冷堆燃料性能驗(yàn)證實(shí)驗(yàn)回路進(jìn)行安全分析，探索超臨界水冷系統(tǒng)的動(dòng)態(tài)和安全特性，提出安全系統(tǒng)設(shè)計(jì)的改進(jìn)方向。本文為超臨界水冷堆燃料性能驗(yàn)證實(shí)驗(yàn)回路的設(shè)計(jì)和安全許可證申請(qǐng)?zhí)峁┝税踩治龉ぞ吆头治鼋Y(jié)果，，對(duì)該實(shí)驗(yàn)回路安全系統(tǒng)的設(shè)計(jì)具有現(xiàn)實(shí)指導(dǎo)意義。
[Abstract]:System safety analysis program (hereinafter referred to as system program) is an important tool for transient and accident safety analysis of reactor system. Under the background of the fourth generation advanced nuclear energy system research and development in the last decade, the international nuclear energy community has carried out a lot of research work on the fourth generation reactor system program. Due to the mature development of the LWR system, some of them have been modified to be suitable for the fourth generation nuclear power system after comprehensive system verification and extensive application. The applicability and reliability of these modified procedures need further study. In this paper, the model development, verification and application of system program for supercritical water cooled reactor and sodium cooled fast reactor in the fourth generation reactor are carried out. Firstly, the applicability of ATHLET program for light water reactor system on these two fourth generation reactor types is evaluated. Based on this, the secondary model of ATHLET program is developed, that is, the relevant calculation model suitable for supercritical water and sodium fluids is developed. The multifluid system program ATHLET-MF (Multi-Fluid) is obtained. Secondly, the modified ATHLET-MF program is preliminarily verified, and the natural convection experiment of France PHENIX sodium cold fast reactor is used to evaluate the program. Finally, The ATHLET-MF program is applied to the safety analysis of the experimental circuit for fuel performance verification of supercritical water-cooled reactor, and some suggestions for improving the design of the safety system of the experimental loop are presented. The main work of the thesis includes: 1. The applicability evaluation and multi-fluid model development of ATHLET program: (1) analyzing the applicability of ATHLET program to supercritical water cooled reactor and sodium cooled fast reactor; (2) increasing physical property calculation, heat transfer, pressure drop and critical flow model of supercritical water; (3) increase the physical properties of sodium, heat transfer, pressure drop model; (4) introduce multi-fluid universal interface, expand the program for supercritical water and sodium into multi-fluid model, obtain the version of ATHLET-MF program. 2. The preliminary verification and evaluation of the multi-fluid model of ATHLET-MF program: (1) the preliminary verification of the expansion function of supercritical water cooled reactor, including the comparison with the theoretical calculation and the program verification; (2) the evaluation of the expansion function of the sodium cooled fast reactor, That is, PHENIX reactor natural convection experimental verification. The results show that: ATHLET-MF program has good applicability and reliability for transient simulation of supercritical water cooled reactor and sodium cooled fast reactor system. The application of the multi-fluid model of ATHLET-MF program, that is, the safety analysis of the experimental circuit for fuel performance verification of supercritical water-cooled reactor: (1) five kinds of design benchmark accidents are calculated and analyzed, including coolant loss, main pump clamp shaft, test circuit power outage. The coolant channel blockage and the coolant side flow accident caused by the rupture of the internal components of the pressure pipe; (2) the improvement measures for the defects of the design of the initial safety system are put forward: two additional safety signals are added, A long-term waste heat discharge scheme is proposed and feasibility analysis is carried out. (3) sensitivity analysis of some design parameters of the safety system is carried out and some suggestions on the control signals and equipment parameters of the safety system are put forward. The purpose of this paper is to develop and verify the system safety analysis tools suitable for supercritical water cooled reactor and sodium cooled fast reactor. The purpose of this paper is to carry out safety analysis for the fuel performance verification experimental loop of supercritical water cooled reactor, and to explore the dynamic and safety characteristics of supercritical water cooling system. The improvement direction of safety system design is put forward. This paper provides a safety analysis tool and analysis results for the design of the experimental loop for fuel performance verification and the application for a safety license for the supercritical water cooled reactor. It is of practical significance for the design of the safety system of the experimental circuit.
【學(xué)位授予單位】：上海交通大學(xué)
【學(xué)位級(jí)別】：博士
【學(xué)位授予年份】：2013
【分類(lèi)號(hào)】：TL364.1
本文編號(hào)：2238288
[Abstract]:System safety analysis program (hereinafter referred to as system program) is an important tool for transient and accident safety analysis of reactor system. Under the background of the fourth generation advanced nuclear energy system research and development in the last decade, the international nuclear energy community has carried out a lot of research work on the fourth generation reactor system program. Due to the mature development of the LWR system, some of them have been modified to be suitable for the fourth generation nuclear power system after comprehensive system verification and extensive application. The applicability and reliability of these modified procedures need further study. In this paper, the model development, verification and application of system program for supercritical water cooled reactor and sodium cooled fast reactor in the fourth generation reactor are carried out. Firstly, the applicability of ATHLET program for light water reactor system on these two fourth generation reactor types is evaluated. Based on this, the secondary model of ATHLET program is developed, that is, the relevant calculation model suitable for supercritical water and sodium fluids is developed. The multifluid system program ATHLET-MF (Multi-Fluid) is obtained. Secondly, the modified ATHLET-MF program is preliminarily verified, and the natural convection experiment of France PHENIX sodium cold fast reactor is used to evaluate the program. Finally, The ATHLET-MF program is applied to the safety analysis of the experimental circuit for fuel performance verification of supercritical water-cooled reactor, and some suggestions for improving the design of the safety system of the experimental loop are presented. The main work of the thesis includes: 1. The applicability evaluation and multi-fluid model development of ATHLET program: (1) analyzing the applicability of ATHLET program to supercritical water cooled reactor and sodium cooled fast reactor; (2) increasing physical property calculation, heat transfer, pressure drop and critical flow model of supercritical water; (3) increase the physical properties of sodium, heat transfer, pressure drop model; (4) introduce multi-fluid universal interface, expand the program for supercritical water and sodium into multi-fluid model, obtain the version of ATHLET-MF program. 2. The preliminary verification and evaluation of the multi-fluid model of ATHLET-MF program: (1) the preliminary verification of the expansion function of supercritical water cooled reactor, including the comparison with the theoretical calculation and the program verification; (2) the evaluation of the expansion function of the sodium cooled fast reactor, That is, PHENIX reactor natural convection experimental verification. The results show that: ATHLET-MF program has good applicability and reliability for transient simulation of supercritical water cooled reactor and sodium cooled fast reactor system. The application of the multi-fluid model of ATHLET-MF program, that is, the safety analysis of the experimental circuit for fuel performance verification of supercritical water-cooled reactor: (1) five kinds of design benchmark accidents are calculated and analyzed, including coolant loss, main pump clamp shaft, test circuit power outage. The coolant channel blockage and the coolant side flow accident caused by the rupture of the internal components of the pressure pipe; (2) the improvement measures for the defects of the design of the initial safety system are put forward: two additional safety signals are added, A long-term waste heat discharge scheme is proposed and feasibility analysis is carried out. (3) sensitivity analysis of some design parameters of the safety system is carried out and some suggestions on the control signals and equipment parameters of the safety system are put forward. The purpose of this paper is to develop and verify the system safety analysis tools suitable for supercritical water cooled reactor and sodium cooled fast reactor. The purpose of this paper is to carry out safety analysis for the fuel performance verification experimental loop of supercritical water cooled reactor, and to explore the dynamic and safety characteristics of supercritical water cooling system. The improvement direction of safety system design is put forward. This paper provides a safety analysis tool and analysis results for the design of the experimental loop for fuel performance verification and the application for a safety license for the supercritical water cooled reactor. It is of practical significance for the design of the safety system of the experimental circuit.
【學(xué)位授予單位】：上海交通大學(xué)
【學(xué)位級(jí)別】：博士
【學(xué)位授予年份】：2013
【分類(lèi)號(hào)】：TL364.1
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