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.     

Nonlinear analysis of nuclear coupled density wave instability in time domain for a boiling water reactor core undergoing core-wide and regional modes of oscillations

The objective of the paper is to develop a nuclear coupled thermal-hydraulic model in order to simulate core-wide (in-phase) and regional (out-of-phase) stability analysis in time domain within the limitation of desktop research facility for a boiling water reactor subjected to operational transients. The integrated numerical tool, which is a combination of thermal-hydraulic, neutronic and fuel heat conduction models, is used to analyze a complete boiling water reactor core taking into account the strong nonlinear coupling between the core neutron dynamics and primary circuit thermal-hydraulics via the void-temperature reactivity feedback effects. The integrated model is validated against standard benchmark and published results. Finally, the model is used for various parametric studies and a number of numerical simulations are carried out to investigate core-wide and regional instabilities of the boiling water reactor core with and without the neutronic feedback effects. Results show that the inclusion of neutronic feedback effects has an adverse effect on boiling water reactor core by augmenting the instability at lower power for same inlet subcooling during core-wide mode of oscillations, whereas the instability is being suppressed during regional mode of oscillations in presence of the neutronic feedback. Dominance of core-wide instability over regional mode of oscillations is established for the present case of simulations which indicates that the preclusion of the former will automatically prevent the latter at the existing working condition.     

A feasibility assessment of the use of nanofluids to enhance the in-vessel retention capability in light-water reactors

Nanofluids, colloidal dispersions of nanoparticles, exhibit a substantially higher critical heat flux (CHF) compared to water. As such, they could be used to enhance the in-vessel retention (IVR) capability in the severe accident management strategy implemented by certain light-water reactors. It is envisioned that, at normal operating conditions, the nanofluid would be stored in dedicated storage tanks, which, upon actuation, would discharge into the reactor cavity through injection lines. The design of the injection system was explored with risk-informed analyses and computational fluid dynamics. It was determined that the system has a reasonably low failure probability, and that, once injected, the nanofluid would be delivered effectively to the reactor vessel surface within seconds. It was also shown analytically that the increase in decay power removal through the vessel using a nanofluid is about 40%, which could be exploited to provide a higher IVR safety margin or, for a given margin, to enable IVR at higher core power. Finally, the colloidal stability of a candidate alumina-based nanofluid in an IVR environment was experimentally investigated, and it was found that this nanofluid would be stable against dilution, exposure to gamma radiation, and mixing with boric acid and lithium hydroxide, but not tri-sodium phosphate.     

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.     

压水堆-回路管道的铸造工艺及其国产化

工艺评定表明,1 000 Mw压水堆核电厂(CPR1000)原选用的主管道铸件Z3CN20-09M(法国牌号)不锈钢的化学成分符合RCC-M采购技术规范,但力学性能并不能完全满足压水堆核岛机械设备设计和建造规范(RCC-M)的要求.本文从金属学角度分析了Z3CN20-09M不锈钢抗蚀性特点和力学性能强化机理,确立了主管道铸件冶炼化学成份的内控标准,使CPR1000核电厂核岛主管道铸件(以下简称主管道铸件)的工艺评定在保持抗蚀性和可焊性特点前提下,各项力学性能指标均满足RCC-M标准,且有较大的裕度,离散度小,质量稳定,综合性能达到领先水平.     

国产快堆包壳材料CW316(Ti)SS高温强度下降的微观机理分析

冷加工316(Ti)不锈钢CW 316(Ti)SS是我国首选的快堆包壳材料,国产材料的常规力学性能与国外数据相当,但高温蠕变和高温持久强度数据却较低.本项研究主要是通过观察、比较国产快堆包壳材料和俄罗斯快堆包壳材料在高温下微观结构的变化情况,并结合对国产材料高温持久断裂试验样品的断口形貌观察结果,分析得出:国产材料长时高温力学性能下降的主要原因是沿晶界的σ相析出.     

AP1000与M310堆型余热排出系统的差异分析

简要介绍了美国西屋公司推出的三代堆型AP1000中正常余热排出系统(RNS)和M310堆型余热排出系统(RRA)的设计特点;分析了余热排出系统在这2种堆型中的主要差异.通过对比这2种堆型中余热排出系统的比较,从工艺系统角度对M310堆型的RRA系统进行局部改进,提高了系统的可靠性和安全性.     

Chemical heat-pump system working at exhaust temperatures of high-temperature-gas-cooled reactor

A chemical heat-pump system using two hydrogen-absorbing alloys is proposed to utilize heat exhausted from a high-temperature source such as high-temperature-gas-cooled reactor, HTGR, which is designed to produce H₂ more efficiently. The overall system proposed here consists of HTGR, He gas turbines, chemical heat pumps and reaction vessels corresponding to the three-step decomposition reactions comprising the I–S process. A fundamental research is performed experimentally on heat generation in a single bed packed with a hydrogen-absorbing alloy that works at the H₂ production temperature. The hydrogen-absorbing alloy of Zr(V_1−xFe_x)₂ is selected as a material that has a proper plateau pressure for the heat-pump system operated between the input and output temperatures of HTGR. Temperature jump due to heat generated when the alloy absorbs H₂ proves that the alloy–H₂ system can heat up the exhaust gas even at 600 °C without any external mechanical force.     

西安脉冲堆~3He快中子夹心谱仪的方案设计

本文介绍了3He快中子夹心谱仪的气体填充、探测器、三通道数字符合和数据分析处理系统,描述了探测器系统的理论分析及实验标定方法,提出了谱仪符合本底计数的来源,绐出了扣除几种符合本底的理论和实验方法,研究结果表明本工作可指导3He快中子夹心谱仪的深入研究.     

池式快堆三维热工流体力学设计研究

随着计算机软硬件技术的发展,三维数值分析技术已经成为池式快堆堆芯和钠池热工设计和计算分析的重要组成部分,并在其中发挥着不可替代的作用.通过对池式快堆几个典型热工现象的分析,展示了我国第一座池式快堆(中国实验快堆)热工设计和安全分析中所拥有的设计手段和工具,总结了三维数值分析技术在快堆工程中的应用,并指出了其对今后快堆热工设计的重要意义.