scholarly journals R&D on a Nonlinear Dynamics Analysis Code for the Drop Time of the Control Rod

2017 ◽  
Vol 2017 ◽  
pp. 1-6
Author(s):  
Daogang Lu ◽  
Yuanpeng Wang ◽  
Qingyu Xie ◽  
Huimin Zhang ◽  
Muhammed Ali
Keyword(s):  
Nonlinear Dynamics ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Response Analysis ◽  
Safe Operation ◽  
Control Rod ◽  
Long Time ◽  
Drop Time ◽  
Dynamics Response

Whether the control rod can drop down in time is one of the important guarantees for the safe operation of the nuclear power plant. The drop-down process of the control rod is very complicated. For a long time, the researchers have done a lot of work on that, but it is hard to consider all the nonlinear factors. This paper considers the main factors together. Based on the theoretical analysis, we developed the nonlinear dynamics response analysis software for the nuclear power plant, which can be used to calculate the rod’s drop-down time. Compared with the results of the experiments, the software we developed proves to be applicable and reliable.

Seismic Response Analysis of Reactor Coolant Pump in Nuclear Power Plant

2008 ◽  
Author(s):  
Zhaohui Ren ◽  
Hui Ma ◽  
He Li ◽  
Guiqiu Song ◽  
Wenjian Zhou
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Structural Design ◽  
Nuclear Power ◽  
Response Spectrum ◽  
Response Analysis ◽  
Response Spectrum Analysis ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Modes Of Vibration

The reactor coolant pump in nuclear power plant is the only revolving equipment in the nuclear power plant. Its functional stability will directly affect the security of nuclear power plant. The coolant pump of a very nuclear plant is examined by using response spectrum analysis to analysis dynamic characteristics and responses aiming at finding the natural frequencies of vibration, modes of vibration and seismic responses, and any possible step which may cause damage of the whole system. The favorable spectrum and unfavorable one are investigated as well. The paper focuses on avoiding the detrimental effects caused by earthquakes, therefore may lay down a theoretical foundation for structural design and installation.

Lessons From an Operational Small Leakage at Tsuruga Nuclear Power Station

2015 ◽  
Author(s):  
Hiroyuki Kobayashi ◽  
Osamu Urabe ◽  
Takushi Fujino
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power Station ◽  
Nuclear Power ◽  
The Other ◽  
Leak Test ◽  
Power Station ◽  
Long Time ◽  
Short Time ◽  
Repair Techniques

Operational small leakage is occasionally observed in a nuclear power plant, and the leak forces an operator to decide whether to shut down the plant or not. Even if the leakage is just a little, it might draw the considerable attention in the society, so that the operator sometimes gets into the situation to judge more severely than technical judgment. Furthermore, at the time of plant restart and the system leak test just after maintenance, even the operator doesn’t accept any leakage considering the long management for the leakage up to the next outage. On the other hand, once the operator shut down the plant, it sometimes takes long time to restart again because of the difficulty to obtain new pipes and valves in short time. The temporary repair techniques referred to the JSME code might be able to be applied to maintain the plant operation, however some difficulties exist in a practical process. One of the authors has faced with many cases in which the operational small leakage had to be dealt at Tsuruga nuclear power station. This paper shows some cases of them and discusses lessons which are related to the codes and standards.

Accident Analysis of Fukushima Daiichi Nuclear Power Plant Unit 2 by the SAMPSON Severe Accident Code

2014 ◽  
Author(s):  
Atsuo Takahashi ◽  
Marco Pellegrini ◽  
Hideo Mizouchi ◽  
Hiroaki Suzuki ◽  
Masanori Naitoh
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Accident Analysis ◽  
Unknown Boundary ◽  
Long Time ◽  
Fukushima Daiichi ◽  
Nuclear Power Plant Unit

The transient process of the accident at the Fukushima Daiichi Nuclear Power Plant Unit 2 was analyzed by the severe accident analysis code, SAMPSON. One of the characteristic phenomena in Unit 2 is that the reactor core isolation cooling system (RCIC) worked for an unexpectedly long time (about 70 h) without batteries and consequently core damage was delayed when compared to Units 1 and 3. The mechanism of how the RCIC worked such a long time is thought to be due to balance between injected water from the RCIC pump and the supplied mixture of steam and water sent to the RCIC turbine. To confirm the RCIC working conditions and reproduce the measured plant properties, such as pressure and water level in the pressure vessel, we introduced a two-phase turbine driven pump model into SAMPSON. In the model, mass flow rate of water injected by the RCIC was calculated through turbine efficiency degradation the originated from the mixture of steam and water flowing to the RCIC turbine. To reproduce the drywell pressure, we assumed that the torus room was flooded by the tsunami and heat was removed from the suppression chamber to the sea water. Although uncertainties, mainly regarding behavior of debris, still remain because of unknown boundary conditions, such as alternative water injection by fire trucks, simulation results by SAMPSON agreed well with the measured values for several days after the scram.

scholarly journals Structural Monitoring and Response Analysis of Prestressed Concrete Primary Containment of a Nuclear Power Plant

2014 ◽  
Vol 86 ◽  
pp. 554-559
Author(s):  
Y.T. Praveenchandra ◽  
Raghupati Roy ◽  
N. Madhusudana Rao ◽  
Arvind Shrivastava
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Prestressed Concrete ◽  
Nuclear Power ◽  
Structural Monitoring ◽  
Response Analysis

A New Safety Analysis Method of Control Rod Ejection Accident in PWR NPP Based on the Failure of Causal Relationship

2018 ◽  
Author(s):  
Shiyu Yan ◽  
Hua Liu ◽  
Zhaohui Liu ◽  
Xiaohua Yang ◽  
Meng Li ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Safety Analysis ◽  
Pressurized Water ◽  
Control Rod ◽  
Drive Mechanism ◽  
Cause And Effect ◽  
Simulation Calculation ◽  
Causal Logic

In view of control rod ejection accident of the traditional pressurized water reactor, the safety thought of the design phase is to validate reliability and availability of DCS I&C in the severe accidents. Now the most important and effective means is simulation calculation and analysis. It is applied for the imaginary accident of the nuclear power plant by using computer software. The new safety analysis steps based on the analysis of cause-and-effect logic failure: firstly, the composition and working principle of control rod drive mechanism is analyzed; secondly, a list of factors-the dynamics and structure, environmental reasons, the function of the control rod drive mechanism and status analysis-are all taken into account, the initial cause of failure modes with causal logic analysis is carried out; thirdly, based on cause-and-effect logic failure, the prevention and improvement measures of accidents, the new criterion of design are put forward. The advantages of cause-and-effect logic failure safety analysis: 1.be based on causal logic. 2. the system aspects is added compared with the past method that is only based on simulation calculation and analysis of the hypothetical accident, the accident the transient process of the key security parameters as the acceptance criteria. 3. The verification and audit of the lack of safety design criteria, completeness of design content, sufficiency problem are performed before the simulated calculation and analysis. 4. The coverage of safety analysis is expanded. Some good advices are provided for the design, operation and maintenance of nuclear power plant.

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