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Chen G.,State Nuclear Power Software Development Center | Chen G.,National Energy Key Laboratory of Nuclear Power Software | Xiao S.,State Nuclear Power Software Development Center | Wang H.,State Nuclear Power Software Development Center | And 2 more authors.
SIMULTECH 2015 - 5th International Conference on Simulation and Modeling Methodologies, Technologies and Applications, Proceedings

Nuclear Power is considered to be one of the solutions to fulfil the increasing need of clean energy in China. Making use of this clean energy can help reduce the consumption of fossil energy, which could enhance the surrounding areas by preventing the environment from harmful air pollution. However, the consequence of severe accident of nuclear power is unbearable, such as Chernobyl and Fukushima Accidents. So as to improve the safety of nuclear power, the severe accident shall be managed in case to reduce the negative impacts to the environment and people health. This paper introduces the Severe Accident Management and Emergency Response System (SAMERS), which aims to help the operators and technicians deal with the severe accidents. Especially, the development of Computerized Severe Accident Management Guidelines (CSAMG) is described in detail, which is a module of SAMERS. CSAMG is based on the AP1000 severe accident management guidelines, which could enhance the operator performance during severe accident. Source

Dong Z.,North China Electrical Power University | Wu J.,North China Electrical Power University | Ma X.,North China Electrical Power University | Ma X.,Purdue University | And 4 more authors.
Annals of Nuclear Energy

The WIMS-D library based on WIMS 69 or XMAS 172 energy group structures is widely used in thermal reactor research. Otherwise, the resonance overlap effect is not taken into account in the two energy group structure, which limits the accuracy of resonance treatment. The SHEM 281 group structure is designed by the French to avoid the resonance overlap effect. In this study, a new WIMS-D library with SHEM 281 mesh is developed by using the NJOY nuclear data processing system based on the latest Evaluated Nuclear Data Library ENDF/B-VII.1. The parameters such as the thermal cut-off energy and lambda factor that depend on group structure are discussed. The lambda factor is calculated by Neutron Resonance Spectrum Calculation System and the effect of this factor is analyzed. The new library is verified through the analysis of various criticality benchmarks by using DRAGON code. The values of multiplication factor are consistent with the experiment data and the results also are improved in comparison with other WIMS libraries. © 2016 Elsevier Ltd. All Rights Reserver. Source

Wang L.,Xian Jiaotong University | Liu L.,Nuclear State Power Software Develop Center | Liu L.,National Energy Key Laboratory of Nuclear Power Software | Tian W.,Xian Jiaotong University | And 5 more authors.
Hedongli Gongcheng/Nuclear Power Engineering

In this paper, we use CFD software ANSYS CFX to simulate the separation phenomena of two-phase (air-water) flowing through branch pipe when the flow pattern in the horizontal pipe is mainly bubble flow for T-junction of AP1000 consisting of ADS-4 and primary loop. The separation ratio, phase profile, pressure profile and velocity profile are obtained, and the effect of different inlet volume fraction and bubble size on the characteristics of separation is also obtained. The results indicate that the phase separation phenomenon at the T-junction position is significant, separation ratio will decrease with an increase of the inlet volume fraction, and there exists a bubble size that makes the phase separation effect most significant for a particular pipe. Source

Li X.,North China Electrical Power University | Li X.,Beijing Key Laboratory of Passive Safety Technology for Nuclear Energy | Li N.,North China Electrical Power University | Li N.,Beijing Key Laboratory of Passive Safety Technology for Nuclear Energy | And 8 more authors.
Annals of Nuclear Energy

The passive residual heat removal system (PRHR) plays an important role in AP1000 passive core cooling system during non-LOCA accidents. The temperature of the water in the in-containment refueling water storage tank (IRWST) will increase while the decay heat is removed through the C-tube heat exchanger under accident conditions, resulting in the change of heat transfer mechanism. Therefore, it is essential to study the relevant pool boiling phenomenon in IRWST by means of experiment. To ensure to capture the major ones from a large amount of influence factors associated with pool boiling in IRWST, the phenomena identification and ranking table (PIRT) was analyzed in this paper. The PRHR system was firstly divided into different modules and components based on the PIRT flow chart, and then the general factors associated with local heat transfer were discussed under different pool boiling stages. Finally, the importance levels of different parameters on heat transfer process were evaluated, which can provide some reference for future design of new PRHR system. © 2016 Elsevier Ltd. All rights reserved. Source

Zhang Y.,North China Electrical Power University | Lu D.,North China Electrical Power University | Wang Z.,State Nuclear Power Software Development Center | Wang Z.,National Energy Key Laboratory of Nuclear Power Software | And 6 more authors.
Applied Thermal Engineering

In AP1000 plant, the Automatic Depressurization System (ADS) 1–3 stages operate to discharge the high-temperature and high-pressure steam from the Reactor Coolant System (RCS) primary side to the large heat sink tank In-containment Refueling Water Storage Tank (IRWST) in accidental conditions. The key equipment's specific shape and arrangement lead to the complicate flow and heat transfer characteristics in IRWST. In the present work, an overall scaled IRWST&ADS sparger experiment has been built up. The thermocouples matrix, flowmeters, pressure transmitters, heat flux sensors, Particle Image Velocimetry (PIV) technique, and high speed camera are employed for the measurements of the key thermal and flow parameters. The local steam jets condensation phenomena as well as the overall flow and thermal behavior are investigated. The experimental results indicate that the thermal stratification phenomenon is obvious in IRWST. The criteria of Richardson Number and Stratification Number are utilized to predict and evaluate the thermal stratification extent, respectively. An improved ADS arrangement design is further proposed to reduce the thermal stratification. Moreover, the multi-holes lumped “steam condensation column” is modeled with characteristic parameters, then the steam condensation heat transfer coefficient range in chugging condensation process is estimated. The experimental results provide practical engineering application reference for the effective operation of the passive safety system in AP1000 plant. © 2016 Elsevier Ltd Source

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