Thermal hydraulic behaviors during loss of RHR system at mid-loop operation of Chinese 300 MWe PWR NPP

Thermal hydraulic behavior of nuclear power plant (NPP) is analyzed by using mechanistic computer code for loss of residual heat removal (RHR) system during mid-loop operation of Chinese 300 MWe two-loop pressurized water reactor is presented. In the absence of recovery of RHR or other accident management measures, the reactor core will be uncovered for a long term resulting in core heat-up, degradation and relocation to the lower plenum. The effectiveness of available mitigate measures, such as safety injection system, gravity feed from refueling water storage tank (RWST) and steam generator (SG) reflux-condensation, are investigated. Coolant injection is highly effective in halting the accident progression and make the core recovered. The cooling capability of SG reflux-condensation has a relationship with different availabilities of steam generators and decay heat power. 6 days after shutdown, 2SG operation can keep the water level at mid-line of hot leg. 12 days after shutdown, both 2SG operation and 1SG operation can keep the water level at mid-line of hot leg. The analyses also indicate that the cooling mechanism of safety injection system is more effective than gravity feed from RWST and SG reflux-condensation. Through confirming the success criteria of SG reflux-condensation, time windows can be devided. Then, event trees for loss of RHR system under mid-loop operation are built with considering the analysis results and abnormal procedure.     

中国实验快堆主容器母材及其颈部支撑焊接件在高温钠及钠蒸气中的腐蚀

在450℃钠和190℃钠蒸气中分别试验研究了中国实验快堆(CEFR)主容器316-316不锈钢焊接件及其颈部支撑316-5083钢焊接件与高温钠及钠蒸气的相容性特征.试验时间为3 000 h,钠中的氧含量及碳含量分别为＜25 μg/g和＜14 μg/g.试验后对腐蚀速率、微观腐蚀形貌、表面元素含量、力学性能进行了分析测试,结果表明,在本试验条件下,两种焊接件与高温钠及钠蒸气的相容性良好.     

反应堆压力容器出口接管管嘴缺陷断裂力学分析

采用断裂力学分析方法,对大亚湾核电站反应堆压力容器出口接管管嘴上的一些缺陷进行了疲劳裂纹扩展分析和快速断裂力学分析,且依据规范对计算结果进行了评定,结果表明:此缺陷不会影响安全.     

压水堆核电厂严重事故下堆腔注水措施研究

针对百万千瓦级压水堆核电厂,采用一体化严重事故分析工具,对一回路冷段大破口冷却剂丧失（LB-LOCA）始发严重事故下,采取堆腔注水（ERVC）缓解措施的事故进程进行模拟,对该措施缓解堆芯熔化进程、保持压力容器完整性的有效性进行分析验证,并对影响该措施的因素进行研究。分析结果表明,在充足的水源条件下,保证一定的注水速率和水位高度,LB-LOCA始发严重事故下采取堆腔注水的缓解措施可为下封头提供有效的冷却,保持压力容器的完整性。     

2022年探矿工程十大新闻

<正>1习近平回信勉励山东省地矿局第六地质大队全体地质工作者在找矿突破战略行动中发挥更大作用奋力书写“英雄地质队”新篇章中共中央总书记、国家主席、中央军委主席习近平2022年10月2日给山东省地矿局第六地质大队全体地质工作者回信,对他们弘扬优良传统、做好矿产勘查工作提出殷切期望。习近平在回信中表示,建队以来,你们一代代队员跋山涉水,风餐露宿,攻坚克难,取得了丰硕的找矿成果,展现了我国地质工作者的使命担当。     

The Fight against Cancer by Microgravity: The Multicellular Spheroid as a Metastasis Model

Cancer is a disease exhibiting uncontrollable cell growth and spreading to other parts of the organism. It is a heavy, worldwide burden for mankind with high morbidity and mortality. Therefore, groundbreaking research and innovations are necessary. Research in space under microgravity (µg) conditions is a novel approach with the potential to fight cancer and develop future cancer therapies. Space travel is accompanied by adverse effects on our health, and there is a need to counteract these health problems. On the cellular level, studies have shown that real (r-) and simulated (s-) µg impact survival, apoptosis, proliferation, migration, and adhesion as well as the cytoskeleton, the extracellular matrix, focal adhesion, and growth factors in cancer cells. Moreover, the µg-environment induces in vitro 3D tumor models (multicellular spheroids and organoids) with a high potential for preclinical drug targeting, cancer drug development, and studying the processes of cancer progression and metastasis on a molecular level. This review focuses on the effects of r- and s-µg on different types of cells deriving from thyroid, breast, lung, skin, and prostate cancer, as well as tumors of the gastrointestinal tract. In addition, we summarize the current knowledge of the impact of µg on cancerous stem cells. The information demonstrates that µg has become an important new technology for increasing current knowledge of cancer biology.     

核反应堆压力容器可靠性分析

本文介绍了反应堆压力容器可靠性的概率分析方法。其主要内容包括初始裂纹、裂纹扩展、载荷条件和物性参数等的概率分布.     

Coupled Ablation and Heat Transfer Analysis of Lower Head for Reactor Pressure Vessel under Severe Accident Condition

After a reactor core melts accident, the solid wall of the reactor pressure vessel (RPV) will be inevitably eroded by the melting core which contains large density of heat flux. The analysis of the coupled ablation and heat transfer of the lower head for RPV is of great theoretical significance to the effectiveness demonstration of water injection in reactor pit and the confirmation of the residual wall thickness of RPV. In this work, numerical simulations were carried out based on the RPV model of CPR1000 using the CFD software FLUENT 17.2. Based on dynamic mesh model and user-defined function (UDF) redevelopment, a fully coupling calculation model considering the transient ablation and heat conduction of solid wall of RPV, the redistribution of heat flux density in RPV inner wall and the subcooled boiling of RPV outer wall was established. Both two-phase flow pattern in the reactor pit and temperature field of RPV solid wall ablation within 9 000 s were obtained and the minimum residual wall thickness and the occurrence location were determined by analysis. The results show that it is feasible to use dynamic mesh to capture wall ablation. The fully coupling calculation model has certain advantages in analyzing the transient ablation process of RPV under severe accident.     

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.     

反应堆压力容器法兰面沟槽尺寸参数对其金属O形环密封性能的影响

基于二维轴对称非线性弹塑性模型,对反应堆压力容器(RPV)用Inconel 718合金O形环进行了有限元分析,在试验验证的基础上探讨了筒体法兰面安装沟槽深度、沟槽间隙等关键参数对密封性能的影响。结果表明：镀银层对O形环的变形行为无实质影响,在进行有限元分析时可忽略镀银层;沟槽深度和沟槽间隙是影响O形环密封性能的关键参数,所涉及的RPV用金属O形环的最佳有效沟槽深度在(11±0.25) mm范围内、沟槽间隙取0.6 mm左右为宜。此外,O形环的安装和制造要求均极为苛刻,为保证O形环与沟槽间距的均匀性应严格控制好制造和安装等工序尺寸参数,从而提高密封性能。