Shandong Nuclear Power Co.

Yantai, China

Shandong Nuclear Power Co.

Yantai, China
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Wang P.,China Nuclear Power Engineering Co. | Fan H.,China Nuclear Power Engineering Co. | Yan Y.,Shandong Nuclear Power Co.
Zhenkong Kexue yu Jishu Xuebao/Journal of Vacuum Science and Technology | Year: 2017

Here, we experimentally addressed thespecial operation environment of the vacuum pumps, dedicated to the degassing tower at AP1000 nuclear power plant. The influence of the working conditions, including the back pressure, type and flow-rate of the pumped gases, on the operating current of the vacuum pump was investigated. The results show that the current depended strongly on the back pressure, but weakly on the flow-rate. For example, as the back pressure increased, the current slowly increased to the rated-current (below 98 kPa) and rapidly shot up (above 210 kPa) to overload the pump, suggesting an upper limit of the back pressure around 100 kPa; the current slowly increased with an increasing flow-rate. The gas type was found to little affect the current. We suggest that the back pressure should be controlled below the upper limit in the pumping of the degassing tower. © 2017, Journal of Vacuum Science and Technology Publishing House. All right reserved.

Yuan K.,Shanghai JiaoTong University | Qie W.Q.,Shandong Nuclear Power Co. | Tong L.L.,Shanghai JiaoTong University | Cao X.W.,Shanghai JiaoTong University
Progress in Nuclear Energy | Year: 2013

Containment depressurization has been implemented for many nuclear power plants (NPPs) to mitigate the risk of containment overpressurization induced by steam and gases released in LOCA accidents or generated in molten core concrete interaction (MCCI) during severe accidents. Two accident sequences of large break loss of coolant accident (LB-LOCA) and station blackout (SBO) are selected to evaluate the effectiveness of the containment venting strategy for a Chinese 1000 MWe NPP, including the containment pressure behaviors, which are analyzed with the integral safety analyses code for the selected sequences. Different open/close pressures for the venting system are also investigated to evaluate CsI mass fraction released to the environment for different cases with filtered venting or without filtered venting. The analytical results show that when the containment sprays can't be initiated, the depressurization strategy by using the Containment Filtered Venting System (CFVS) can prevent the containment failure and reduce the amount of CsI released to the environment, and if CFVS is closed at higher pressure, the operation interval is smaller and the radioactive released to the environment is less, and if CFVS open pressure is increased, the radioactive released to the environment can be delayed. Considering the risk of high pressure core melt sequence, RCS depressurization makes the CFVS to be initiated 7 h earlier than the base case to initiate the containment venting due to more coolant flowing into the containment. © 2013 Elsevier Ltd. All rights reserved.

Zou J.,Shanghai JiaoTong University | Li Q.,Shandong Nuclear Power Co. | Tong L.L.,Shanghai JiaoTong University | Cao X.W.,Shanghai JiaoTong University
Progress in Nuclear Energy | Year: 2014

Advanced passive PWR relies on passive safety systems to provide core cooling capacity and deal with design basis accidents and beyond design basis accidents. However, the passive safety system is lack of practical operating experience and their performance is heavily influenced by other systems. The cooling capacity of passive residual heat removal system (PRHR), which is designed to remove decay heat when normal heat removal approach is not available, requires specific assessment during different accidents. In this study, a detail model of advanced passive PWR, including Reactor Coolant System (RCS), simplified secondary side and Engineered Safety Features (ESF), has been built using mechanism accident analysis code. The plant transient has been simulated, and cooling capacity of PRHR been analyzed during loss of normal feedwater and main feedwater line rupture. Conservative assumptions were made specially based on different accident scenarios and one of the two fail-open valves arranged in parallel at the PRHR heat exchanger (HX) outlet line was assumed not open, as the worst single failure. The progress of the two accident sequence is calculated and the thermalhydraulic behavior of RCS is investigated and the main transient parameters are obtained, including primary side pressure, steam generator pressure, pressurizer water level. The cooling power and system response are calculated. The results show that PRHR, with CMT injection, can remove the decay heat from RCS to IRWST, keeping the pressures of RCS and steam generators remaining below 110 percent of the design values and the pressurizer overfilling is prevented. Sensitivity study has been performed to study the system resistance effects on the capacity of PRHR, which shows that increase in system resistance coefficient reduces the cooling capacity of PRHR. © 2013 Elsevier Ltd. All rights reserved.

Zou J.,Harbin Institute of Technology | Zhao M.,Harbin Institute of Technology | Wang Q.,Harbin Institute of Technology | Wu G.,Shandong Nuclear Power Co.
IEEE Transactions on Industrial Electronics | Year: 2012

The magnetic field computation and performance characteristics analysis for a tubular transverse flux machine (TFM) with permanent-magnet excitation have been discussed. The following issues have been emphasized: construction of the tubular TFM, magnetic flux distribution, winding back electromotive force, winding inductance, and thrust force. The topology of such a machine requires a 3-D finite-element method to accurately predict the machine performance. The simplified computation model is proposed in order to save the computing time. The experimental setup has been developed. The calculated results and the measured results are in good agreement. Thus, the validity of the simplified computation model is indirectly proved. The thrust-force capabilities of the tubular TFM are compared with other topologies of linear machines. The technology of the tubular machine could be enriched. © 2006 IEEE.

Zhu G.,Nuclear Power Institute of China | Tian X.,Nuclear Power Institute of China | Wei W.,Shandong Nuclear Power Co.
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2015

The beltline region and inlet nozzle of the reactor pressure vessel (RPV) is investigated in this paper. Fracture mechanics finite element model is built, the transient temperature filed and stress filed are analyzed using the detailed thermal engineering analysis result of typical event transients as the input conditions. Combining with the irradiation embrittlement assessment result, the structure integrity of RPV under events is analyzed and assessed adopting the analysis method for the deterministic fracture mechanics. The analysis result shows that brittle rupture would not happen in the interesting region of RPV during the life time of forty years, but attention should be paid to the transients with great changing rate of coolant temperature. ©, 2015, Editorial Office of Nuclear Power Engineering. All right reserved.

Zhu M.,Harbin Engineering University | Tian R.,Harbin Engineering University | Xing Z.,Shandong Nuclear Power Company | Liu Y.,Harbin Engineering University
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2015

Wire-mesh mist eliminator is widely applied in sea water desalination apparatus because of the high economy and efficient removal of liquid droplets from vapor. Open literature on the research of demister performances with the wet working condition at atmosphere is limited. With this article, an experimental test apparatus and device were designed and established. The layered type eliminators made of stainless steel wires with the size of 0.1×0.4mm were 80mm in diameter. The study on steam vapor pressure drop of the wire mesh demister as a function of operating condition and design parameters was carried out. These factors include vapor velocity (2-10m/s), pad thickness (50-190mm), layer spacing (0.5-2mm). Experiments working medium was vapor-liquid at atmospheric pressure. The total pressure drop were found to increase with the increasing of vapor velocity, pad thickness and packing density, the specific pressure drop of each layer mesh pad declined until the vapor velocity increased and reached certain value and then increased with a further growth in vapor velocity. The optimum layer spacing of demister varies varies from 1-1.5mm under the experimental conditions. Copyright © 2015 by JSME.

Zhang H.,Beijing Institute of Petrochemical Technology | Zhuang Q.,Shandong Nuclear Power Co. | Zhang H.,CRRC Qingdao Sifang Co.
Hanjie Xuebao/Transactions of the China Welding Institution | Year: 2016

The intergranular corrosion behavior of friction stir welded 2219 aluminium alloy was investigated. Microstructure, micro-hardness, corrosion morphology and corrosion depth were studied to analyze the difference between BM and WNZ, and intergranular corrosion mechanism of FSW joint was preliminary discussed. The results show that WNZ consists of fine equiaxed grains, and the grain size on top surface is slightly bigger than that on root surface. The highest microhardness is located in the BM while the lowest in the WNZ on root surface. The corrosion resistance of WNZ is much superior than that of BM, and WNZ on top surface is slightly superior than that on root surface. The maximum corrosion depth in BM is 145.9 μm while the maximum corrosion depths in WNZ on top surface and root surface are 46.3 μm and 84.1μm respectively. © 2016, Editorial Board of Transactions of the China Welding Institution, Magazine Agency Welding. All right reserved.

Zou W.-X.,Shandong Nuclear Power Co. | Zou J.-M.,Shandong Nuclear Power Co.
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2014

The paper introduces the application status of the Reliability-Centered Maintenance(RCM) technology in Haiyang nuclear power plant, and analyzes the RCM application in the future, including the optimization of equipment reliability classification, development of performance monitoring program and preventive maintenance program, optimization of preventive maintenance program, risk significant components management and optimization of spare parts. The application of RCM can help utility to develop integral equipment maintenance strategy, improve the equipment reliability and availability, and reduce maintenance costs.

Liu Y.,Shandong Nuclear Power Co. | Zhang H.,Shandong Nuclear Power Co.
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2014

AP1000 as the third generation of nuclear power units, requires its operation water to be below 50 ppm of TOC as described in its Chemistry Manual. TOC contained in Demineralized Water (DW) is mainly the residual of the process of DW raw water pre-treatment. Another source of TOC is owing to decomposition of the organic materials in the system. When exposed to high temperature or radiation, TOC yields organic acid which will corrode metallic materials. TOC contained in DW can be effectively removed by operations of mixed condensing, reverse osmosis (RO), ultra-violet ray radiation, and etc.

Xue Y.,Shandong Nuclear Power Company | Xue T.,China Electric Power Research Institute
World Information on Earthquake Engineering | Year: 2016

The nonlinear finite element analysis software ABAQUS is adopted to analyze the seismic performance and compared with the original structure. The analysis results show that under strong earthquakes, the structural interlayer displacement angle can satisfy the requirements of specification, its seismic performance is significantly improved and can continue to use after the quake after setting viscous dampers. Finally, the research provides a strong theoretical basis for the application of the new type of viscous dampers in the main machine hall in conventional island. © 2016, Science Press. All right reserved.

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