A Study on the Formation and Removing Mechanisms of a Gas Hydrate Slug in Gas Pipelines

Most of the existing hydrate prediction models cannot describe the formation process of hydrate slug in gas pipelines. The authors developed the physical model of hydrate formation state in gas pipelines, and the corresponding mathematical model is proposed by utilizing the theories of heat transfer, multiphase flow, and phase equilibrium. The formation process of hydrate slug is described by simulation, and the removing mechanism of hydrate slug using depressurization is analyzed. The change of related parameters is obtained. which can effectively provide theoretical guidance for preventing and removing the hydrate slug.     

石油磺酸盐降压增注试验研究

针对低渗区块注水困难的状况,在室内开展了石油磺酸盐的溶解性、表面张力变化及岩心试验,检验石油磺酸盐的降压增注效果,并对添加石油磺酸盐是否对注水井造成损害进行了结垢趋势和腐蚀试验。结果表明低浓度石油磺酸盐可降低污水的表面张力,具有降压增注的功效,并且可以起到一定的缓蚀作用。     

盐水液滴降压蒸发析盐过程传热传质特性

针对单个盐水(NaCl溶液)液滴在降压环境下蒸发析盐的传热传质过程建立了数学模型。模型考虑了多孔盐壳在液滴表面的形成过程,降压过程引起的气流运动,液核通过多孔介质的传质扩散,以及液滴表面的蒸发换热和对流换热。将实验数据与计算结果对比,验证了模型的有效性。通过模型计算获得了液滴表面温度及液滴质量随时间的变化。结果表明盐水液滴在降压环境下蒸发析盐过程的温度变化分为4个阶段:温度骤降阶段、温度回升阶段、平衡温度阶段和温度上升阶段。平衡温度阶段,盐壳界面运动较慢,随蒸发进行,液核尺寸逐渐减小,盐壳界面运动速度加快。理论分析了环境压力对盐水液滴蒸发析盐过程的影响,环境压力越低,平衡温度越低,盐分完全析出时间越短。     

超临界压力状态氟利昂泄压实验研究

超临界状态的流体从超临界状态向次临界状态的泄压过程,对超临界水冷堆（SCWR）的安全有非常重要的影响。本文通过相似性原理及模化方法,利用临界点较低的氟利昂（R134a）代替水对超临界水冷堆的泄压过程进行了实验研究。泄压过程分别从容器顶部和底部进行,实验集中观测了超临界态流体泄压过程中压力和温度的瞬态变化。实验分析了流体初始状态、泄压孔直径对泄压过程的影响。在等熵泄压过程中观测到工质从超临界状态进入次临界状态时,两相开始产生的压力和温度主要取决于初始状态及泄压的速率。在等熵泄压过程实验中未观测到容器内工质中形成明显的轴向温度分布。     

水力控制棒驱动系统失压工况的实验研究

利用200MW低温核供热堆水力控制棒驱动系统的1：1实验台架模拟系统失压况，进行控制棒步升，步降，开阀落棒及关泵开阀落棒实验，并与正常工况下的提棒，落棒的实验结果进行比较，实验结果表明：在系统失压工况下，控制棒能正常提棒，落棒系统失压工况下的弹棒，系统压力与压力壳压力具有瞬时跟随特性，未出现控制棒弹棒事故，对实验的失压速率和事故分析得到的失压速率进行了比较，验证了系统具有良好的安全性和可靠性。     

PWR Cold-Leg Small-Break LOCA with Total HPI Failure

Cold-leg small-break loss-of-coolant accident (LOCA) tests were performed at the ROSA-IV Large Scale Test Facility (LSTF), a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). The tests were conducted for break areas ranging 0.5–10% of the scaled cold leg area, and simulated hypothetical total failure of the high pressure injection (HPI) system. One of the tests, conducted with 1% break area, included an intentional depressurization of the primary system that was initiated after the onset of core dryout. A simple prediction model is proposed for prediction of times of major events, namely, loop seal clearing, core dryout, accumulator (ACC) injection and actuation of low pressure injection (LPI) system. Test data and model calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of approximately 5% or more, and might be insufficient for intermediate break areas to maintain adequate core cooling. It is also shown that there might be possibility of core dryout after ACC injection and before LPI injection for break areas less than approximately 2.5%.     

Detailed Evaluation of RCS Boundary Rupture during High-Pressure Severe Accident Sequences

A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture.     

Analysis of different containment models for IRIS small break LOCA, using GOTHIC and RELAP5 codes

Advanced nuclear water reactors rely on containment behaviour in realization of some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts, like the AP600/1000 ones.In this work the international reactor innovative and secure (IRIS) was taken as reference, and the relevant condensation phenomena involved within its containment were investigated with different computational tools. In particular, IRIS containment response to a small break LOCA (SBLOCA) was calculated with GOTHIC and RELAP5 codes. A simplified model of IRIS containment drywell was implemented with RELAP5 according to a sliced approach, based on the two-pipe-with-junction concept, while it was addressed with GOTHIC using several modelling options, regarding both heat transfer correlations and volume and thermal structure nodalization. The influence on containment behaviour prediction was investigated in terms of drywell temperature and pressure response, heat transfer coefficient (HTC) and steam volume fraction distribution, and internal recirculating mass flow rate. The objective of the paper is to preliminarily compare the capability of the two codes in modelling of the same postulated accident, thus to check the results obtained with RELAP5, when applied in a situation not covered by its validation matrix (comprising SBLOCA and to some extent LBLOCA transients, but not explicitly the modelling of large dry containment volumes).The option to include or not droplets in fluid mass flow discharged to the containment was the most influencing parameter for GOTHIC simulations. Despite some drawbacks, due, e.g. to a marked overestimation of internal natural recirculation, RELAP5 confirmed its capability to satisfactorily model the basic processes in IRIS containment following SBLOCA.     

Two types of a passive safety containment for a near future BWR with active and passive safety systems

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.     

二级PSA中人员可靠性分析方法研究

在概率安全分析(PSA)中,人员可靠性分析(HRA)是必不可少的组成部分。国内在一级PSA中的HRA做了大量的研究工作,已有良好的基础和工程实践,但由于核电厂严重事故下人员响应的复杂性,有关二级PSA的HRA还处于摸索阶段。通过研究二级PSA中人员响应特点,调研国内外在二级PSA中采用的HRA方法,最后以我国某三代压水堆核电厂严重事故下一回路快速卸压为例,采用THERP、HCR+THERP以及SPAR-H三种方法,分别进行了HRA,并给出相应的结论和建议。