钠冷快堆多模块蒸汽发生器大泄漏钠水反应事故保护系统关键参数敏感性分析

钠冷快堆采用钠-钠-水三回路设计,当发生传热管破裂后,引起的大泄漏钠水反应事故将威胁二回路的完整性和安全性,设置的保护系统要能够有效保证二回路的完整性。本文以钠冷快堆二回路和多模块蒸汽发生器保护系统为研究对象,建立了大泄漏钠水反应模型,利用钠水反应实验结果对模型进行了验证。模拟了3根传热管发生双端断裂（3-Double-Ended Guillotine,3-DEG）的大泄漏钠水反应过程,分析了二回路的完整性和保护系统的响应。选取保护系统5种关键参数进行敏感性分析,计算其对二回路最高压力和保护系统的影响,结果表明：较小的液相爆破片爆破压力和爆破片爆破延迟时间、较短的泄放管长度以及液相爆破片放置在下腔室,将更有利于保护系统响应和二回路的完整性。     

快堆蒸汽发生器大泄漏钠-水反应计算

采用一维特征线方法建立了快堆蒸汽发生器单管发生双端断裂情况下 ,水从破裂传热管流出的泄漏率计算模型和钠 水反应气泡从球状到柱状的变温绝热生长模型。根据断裂处氢气压力变化 ,并考虑管内流体的压缩性建立了水的泄漏率模型。对分别发生在单相水区 ,单相蒸汽区和两相汽水混合区的大泄漏钠 水反应进行瞬态计算和定性分析。结果说明 ,传热管不同位置发生泄漏对二回路造成的影响不同     

钠冷快堆蒸汽发生器小泄漏事故中氢迁移行为研究

钠冷快堆采用钠-钠-水/蒸汽三回路传热模式,二回路钠与三回路水/蒸汽通过蒸汽发生器实现热交换。蒸汽发生器中传热管的微小破损都可能导致钠水反应。为了有效扼制小泄漏事故的扩展,需要及时发现泄漏的发生。本文建立了钠冷快堆蒸汽发生器小泄漏钠水反应一维计算模型,采用Fortran语言编写了一维分析程序,用于计算小泄漏钠水反应氢气产生、迁移过程,并与参考文献计算结果进行了对比。最后,针对蒸汽发生器一根传热管破损分析了泄漏率、钠温对氢离子和氢气在二回路钠中迁移行为的影响。可为钠冷快堆二回路小泄漏探测系统的布置提供参考。     

大泄漏钠水反应引起压力波传播的研究

快堆蒸发器内管道发生破损、断裂，水／水蒸汽泄漏涌入液钠空间，会产生剧烈的钠水反应事故，瞬时激发很强的压力波，危及管路和部件。为此要求建立一组完整的钠水反应引起压力波传播的数学模型，进行数值计算，定性分析其基本特征。水力实验证明压力波传播数学模型是正确的，实验钠回路的数值计算也说明全组模型是合理的     

中国实验快堆大泄漏反应事故模拟计算分析

大泄漏钠?水反应是钠冷快堆的设计基准事故,可能会导致二回路边界的破坏,导致放射性外泄,研究大泄漏钠水反应事故对反应堆安全分析具有重要的意义.本文建立了中国实验快堆大泄漏反应事故的计算分析模型,计算了反应区压力、反应区温度、二回路管道、重要设备中压力和流量以及缓冲罐气腔体积和缓冲罐钠液位增量等参数变化.将计算结果和中国实...     

快堆钠回路水锤程序开发与应用

研究开发了快堆钠回路水锤分析专用程序WHA。该程序在一维特征线法（MOC）传统的压力波传播数学模型中补充了钠腔－气腔外边界模型，并采用气泡离散模型模拟低压液柱分离中的蒸汽穴的生成与溃灭。程序用FORTRAN90语言对快堆实验钠回路ESPRESSO中由于阀门的快速开启与关闭引起的压力波传播进行了分析计算。计算结果表明：将钠腔－气腔引入水锤压力波传播的数学模型进行程序计算的结果是合理的。     

16 CEFR二回路水锤分析

CEFR以液态钠作为一、二回路冷却剂。二回路蒸汽发生器泄漏发生钠水反应会在二回路系统中引起水锤，水锤事故可以引起管道的甩摆，拉断管道支撑，使管道破裂，甚至直接造成核电厂事故停堆。因此，水锤是CEFR设计和安全分析中的一个重要问题。     

CEFR二回路水锤分析

CEFR以液态钠作为一、二回路冷却剂。二回路蒸汽发生器泄漏发生钠水反应会在二回路系统中引起水锤,水锤事故可以引起管道的甩摆,拉断管道支撑,使管道破裂,甚至直接造成核电厂事故停堆。因此,水锤是CEFR设计和安全分析中的一个重要问题。本工作在清华大学按照CEFR一、二回路系统初步设计参数编制的水锤分析程序的基础上,采用CEFR二回路冷却系统施工设计参数重新进行分析计算。CEFR二回路水锤分析@唐龙 @王彬     

钠水反应试验系统大泄漏钠水反应压力分布计算分析

本文采用一维大泄漏钠水反应分析程序LLEAK,计算和分析了大泄漏钠水反应工况下压力在钠水反应试验系统F204内的分布特性。结果表明:试验系统在水泄漏量为57 g/s、钠循环流量为10 m~3/h和20 m~3/h的工况下,泄漏点压力峰值仅为0.92 MPa,反应器出入口的爆破片均无动作,为保证系统压力处于较低水平,应考虑将爆破片动作整定值从1.0 MPa调低至0.8 MPa;试验系统在水泄漏量为290 g/s时系统压力峰值达到了1.4 MPa,为满足设备安全性,系统全环路的设备应至少能承受2.0 MPa的压力;在较大水泄漏量时,随钠循环流量的增大,将降低钠水反应对缓冲罐压力的影响;在较小水泄漏量时,随钠循环流量的增大,将恶化钠水反应对缓冲罐压力的影响。     

钠水反应试验系统大泄漏钠水反应压力分布计算分析

本文采用一维大泄漏钠水反应分析程序LLEAK,计算和分析了大泄漏钠水反应工况下压力在钠水反应试验系统F204内的分布特性。结果表明：试验系统在水泄漏量为57 g/s、钠循环流量为10 m³/h和20 m³/h的工况下,泄漏点压力峰值仅为0.92 MPa,反应器出入口的爆破片均无动作,为保证系统压力处于较低水平,应考虑将爆破片动作整定值从1.0 MPa调低至0.8 MPa;试验系统在水泄漏量为290 g/s时系统压力峰值达到了1.4 MPa,为满足设备安全性,系统全环路的设备应至少能承受2.0 MPa的压力;在较大水泄漏量时,随钠循环流量的增大,将降低钠水反应对缓冲罐压力的影响;在较小水泄漏量时,随钠循环流量的增大,将恶化钠水反应对缓冲罐压力的影响。     

Design and modification of steam generator safety system of FBR MONJU

MONJU is a prototype fast breeder reactor (FBR) in Japan. The sodium–water reaction in the steam generator (SG) is one of the important safety assessment items for a sodium cooled reactor like MONJU. MONJU is equipped with hydrogen gas detectors for the small water leak detection, gas pressure gauges for the medium leak and sensors of rupture discs for the large leak. As a design basis accident, one tube failure then failure propagation of neighboring three tubes is assessed to verify the structural integrity of the secondary components. A latest evaluation method on the design margin against the overheating tube rupture showed that the present SG system had not an enough margin in the worst case. For improving the margin, it needs to shorten the time of the sodium–water reaction by earlier water leak detection in the SG and sooner water ejection from the SG tubes. Therefore, MONJU is now carrying out the following modification works: (1) addition of steam relief valves, (2) addition of a gas pressure gauge with changing the interlock logic and lowering the trigger level, (3) reducing the opening of the valves on the SG gas flow line to the dump tank because of earlier detection for the pressure rise. After this modification, the design margin of the SG system will be sufficiently improved.     

Analysis of EBR-II secondary sodium network for pressures pulses due to leaks of steam of water into sodium

A systematic procedure is presented for the safety analysis of the secondary sodium system of an LMFBR in the event of a leak of water or steam into sodium in an evaporator or a superheater. Using fracture mechanics, it is shown that the usual assumption of failure initiation by guillotine rupture of one or more water or steam tubes is unrealistic. A model is proposed for the gradual growth of leaks due to the phenomenon of wastage. The pressure rise in the system due to a sodium-water reaction is calculated solving one-dimensional hydrodynamic equations. By comparing results obtained for the proposed gradual growth model with those due to guillotine failure it is shown that the assumption of guillotine failure leads to a significant overestimation of pressures and stresses. Based upon the proposed leak progression model the stresses in the EBR-II secondary sodium system are shown to be within safe limits.     

快堆蒸汽发生器热力参数对泄漏探测系统响应特性的影响

研究建立了水泄漏引起的钠水反应产物在快堆蒸汽发生器和取样支路传输扩散的一维数学模型,分析了蒸汽发生器流量、钠温度和取样支路流量对泄漏探测系统响应特性的影响。模型计算和实验结果表明：蒸汽发生器流量的增加将缩短系统的响应时间,但却降低了蒸汽发生器钠出口处的氢离子浓度,使系统探测水泄漏的灵敏度降低;蒸汽发生器钠温度对系统的响应时间影响不大,钠温升高,OH^-离子的离解速率加快,探测系统的灵敏度提高;增大取样支路流量可改善系统的响应特性。     

Analysis of the Water Hammer in the Secondary Loop of CEFR

The coolant of the primary loop and secondary loop of CEFR is sodium. The reaction of sodium with water during the leakage of SG will induce the water hammer in the secondary loop. The water hammer in the loop would cause the pipeline to swing, the support of the pipeline to snap, the pipeline to break, even the nuclear power station to shutdown directly. The water hammer is an important issue of the design and safe analysis of CEFR.     

Proposed Method of Measuring Thermal Diffusivity of a Small Specimen at Incandescent Temperatures

An evaluation was made of heat transfer tube failure propagation due to sodium-water reaction wastage in a sodium heated steam generator, by comparing an empirically derived wastage equation with leak detector responses.

The experimental data agreed well with the wastage equation even for different values of distance-to-nozzle diameter ratio, though the formula had been based on wastage data obtained for only one given distance. The time taken for failure propagation was estimated for a prototype steam generator, and compared with the responses characteristics of acoustic detectors and level gages. It was found that there exists a range of leak rate between 0.5 and 100g/sec, where the level gage can play a useful role in leak detection. The acoustic detector can be expected to respond more rapidly than the cover gas pressure gage, if noise is kept below ten times the value observed in an experimental facility, SWAT-2.     

An equilibrium and kinetic study of the liquid sodium-hydrogen reaction and its relevance to sodium-water leak detection in LMFBR systems

A detailed study of the thermodynamic and kinetic properties of the liquid sodium-hydrogen system has been made. The technique employed was to measure the rate of absorption to equilibrium of successive hydrogen additions by a vigorously stirred excess of liquid sodium (< 10 ppm dissolved oxygen) in an isothermal, constant volume, stainless steel (316) reaction vessel in the temperature range 610–677 K. The results have been used to estimate the variation of hydrogen pressure with time for a given, steady, water leak rate into the argon gas blanket above the liquid sodium level in a secondary heat exchanger of an LMFBR, and also to calculate solution rates of hydrogen bubbles (formed from a sodiumwater reaction under the liquid sodium surface) as a function of initial bubble radius.     

快堆蒸汽发生器声学泄漏探测技术的研究──Ⅰ . 系统信号特性

根据在实验Ｎａ回路上的实验结果，分析了快堆电站蒸汽发生器中Ｎａ噪声、Ｈ２Ｏ噪声、泵和风机噪声等泄漏信号和背景噪声的时域、频域特性，对泄漏噪声的提取方法提出了设想，为提高泄漏探测系统的灵敏度提供了实验依据。     

Improvement of steam generator tube failure propagation analysis code LEAP for evaluation of overheating rupture

As a part of safety assessment or design of steam generators of sodium-cooled fast reactors, it is necessary to evaluate the water leak rate under sodium–water reaction accident. The computer code called LEAP-II calculating a design basis water leak rate during long-time event progress including self-wastage, target-wastage, wastage-type failure propagation, water leak detection, and water/steam blowdown was developed for the prototype fast reactor in the past studies. In this study, a numerical analysis method to predict occurrence of overheating tube rupture was constructed and integrated into this code to expand its application range. The newly constructed method consists of the elemental analysis models for temperature distribution formed by a reacting jet, water-side thermal hydraulics, heat transfer at the tube wall, temperature and stress of the tube, and failure of the tube. Applicability of the method was investigated through the numerical analysis of the experiment on water vapor discharging into liquid sodium pool under the actual condition of the steam generator. The numerical analysis demonstrated that the method could provide the appropriately conservative result on the overheating-rupture-type failure propagation.     

Small Leak Detection by Measuring Surface Oscillation during Sodium-Water Reaction in Steam Generator

Small leak sodium-water reaction tests were conducted to develop various kinds of leak detectors for the sodium-heated steam generator in FBR. The super-heated steam was injected into sodium in a reaction vessel having a sodium free surface, simulating the steam generator. The level gauge in the reaction vessel generated the most reliable signal among detectors, as long as the leak rates were relatively high. The level gauge signal was estimated to be the sodium surface oscillation caused by hydrogen bubbles produced in sodium-water reaction.

Experimental correlation was derived, predicting the amplitude as a function of leak rate, hydrogen dissolution ratio, bubble rise velocity and other parameters concerned, assuming that the surface oscillation is in proportion to the gas hold-up. The noise amplitude under normal operation without water leak was increased with sodium flow rate and found to be well correlated with Froud number. These two correlations predict that a water leak in a “MONJU” class (300 MWe) steam generator could possibly be detected by level gauges at a leak rate above 2g/sec.