Scaling Evaluation of the Two-Phase Natural Circulation

2012 ◽  
Author(s):  
Han Wang ◽  
Yuquan Li
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Characteristic Time ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Scaling Analysis ◽  
Top Down ◽  
Two Phase ◽  
Height Scale ◽  
Full Pressure

This paper presented the scaling evaluation of the two-phase natural circulation process between an assumed nuclear power plant and three test facilities with full pressure simulation and three different height scales, which were 1:2, 1:3 and 1:4. The Hierarchical Two-Tiered Scaling (H2TS) Methodology was adopted. By top-down scaling analysis, several characteristic time ratios were obtained, and then the calculation method of the scaling distortion were investigated. It has been found that the dominant processes in two-phase natural circulation can be well preserved no matter what the height scale is.

scholarly journals Natural Circulation in the ATUCHA-I PHWR Nuclear Power Plant

2007 ◽  
Vol 2007 ◽  
pp. 1-9
Author(s):  
O. Mazzantini ◽  
J. C. Ferreri ◽  
F. D'Auria ◽  
C. P. Camusso
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heavy Water ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Two Phase ◽  
Integral Test ◽  
Core Mass ◽  
Flow Maps ◽  
Flow Map

A systematic study of natural circulation (NC) in a postulated, varying primary mass inventory scenario at residual power fractions has been performed for a nuclear power plant operating in Argentina. It is a pressurized heavy water reactor, cooled and moderated by heavy water. The analysis seems particularly relevant at present, because a second nuclear power plant (NPP), of similar design and nearly 745 MWe, is now under finalization. NRC-RELAP5/MOD3.3 was the code used to perform the simulations. Results obtained are presented in the form of natural circulation flow maps. The trends obtained fit in the expected limits for integral test facilities representative of PWRs. In addition, the validity of a simplified analysis to scale single and two-phase core flow has been verified. A set of constants has been obtained, which permits predicting NC core mass flow rate (CMFR) for this NPP. Results are partially validated, for single-phase NC flow, using a documented plant transient, showing reasonable agreement. Also, the effect of pressurizer size on the predicted evolution curve in the NC flow map (NCFM) is discussed.

scholarly journals Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

2008 ◽  
Vol 2008 ◽  
pp. 1-11 ◽  
Author(s):  
Avinash J. Gaikwad ◽  
P. K. Vijayan ◽  
Sharad Bhartya ◽  
Kannan Iyer ◽  
Rajesh Kumar ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Transport ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Heat Removal ◽  
Transport Systems ◽  
Critical Examination ◽  
Decay Heat ◽  
Station Blackout

Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP) safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs), like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs) is resorted to mitigate consequences of station blackout (SBO). In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR) system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT), SGs, and PDHRs) under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections). On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

Reliability of PCCS in AP1000 Based on Accident Development

2016 ◽  
Author(s):  
Yu Yu ◽  
Shengfei Wang ◽  
Fenglei Niu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Physical Process ◽  
System Reliability ◽  
Nuclear Power ◽  
Cooling System ◽  
Natural Circulation ◽  
System Structure ◽  
System Failure ◽  
Process Failure

Passive containment cooling system (PCCS) is an important safety-related system in AP1000 nuclear power plant, by which heat produced in reactor is transferred to the heat sink – atmosphere – based on natural circulation, independent of human response or the operation of outside equipments, so the reactor capacity of resisting external hazards (earthquake, flood, etc.) is improved. However since the system operation based on natural circulation, many uncertainty factors such as temperatures of cold and heat sources will affect the system reliability, and physical process failure becomes one of the important contributors to system failure, which is not considered in the active system reliability analysis. That is, the system will lose its function since the natural circulation cannot be established or kept even when the equipments in the system can work well. The function of PCCS in AP1000 is to transfer the heat produced in the containment to the environment and to keep the pressure in the containment below its threshold. After accidents the steam is injected to the containment and can be cooled and condensed when it arrives at the containment wall, then the heat is transferred to the atmosphere through the steel vessel. So the peak value of the pressure is influenced by the steam situation which is injected into the containment and the heat transfer and condensate processes under the accidents. In this paper the dynamic thermal-hydraulic (T-H) model simulating the fluid performance in the containment is established, based on which the system reliability model is built. Here the total pressure in the containment is used as the success criteria. Apparently the system physical process failure may be related to the system working state, the outside conditions, the system structure parameters and so on, and it’s a heavy work to analyze the influences of all the factors, so only the effects of important ones are included in the model. Monte Carlo (MC) simulation is used to evaluate the system reliability, in which the input parameters such as air temperature are sampled based on their probabilistic density distributions. The pressure curves along with the accident development are gained and the system reliabilities under different accidents are gotten as well as the main contributors. The results illustrate that the system physical process failure probabilities are varied under different climate conditions, which result in the system reliability and the main contributors to system failure changing, so the different methods can be taken to improve the system reliability according to the local condition of the nuclear power plant.

Effects of Rolling on Characteristics of System Under Forced Circulation and Natural Circulation

2020 ◽  
Author(s):  
Rong Cai ◽  
Nina Yue ◽  
Hongyu Fang ◽  
Baowen Chen ◽  
Lili Liu ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Natural Circulation ◽  
Thermal Parameters ◽  
Reactor System ◽  
Hydraulic Characteristics ◽  
Forced Circulation ◽  
Rolling Condition

Abstract The marine nuclear power plant operating in the marine environment has complicated motion under the influence of wind and waves. The movement of marine nuclear power plant will affect the thermal-hydraulic characteristics of its nuclear reactor system. Compared with other typical motion conditions, the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system are the most complex. In order to study the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system, a thermal-hydraulic system code for motion conditions named STAC was developed. The STAC code was verified by the experiments conducted in Japan. The effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system under forced circulation and natural circulation are studied with the STAC code. The simulation results show that the thermal parameters of the reactor system under rolling condition fluctuate periodically. The fluctuation period of the thermal parameters of the core is half of the rolling period, and the fluctuation periods of other thermal parameters are the same as the rolling period. The effect of rolling condition on thermal-hydraulic parameters under forced circulation is smaller than that under natural circulation. The fluctuation amplitudes of the thermal parameters increase with the angle amplitude of the rolling condition. There is a rolling period with the smallest fluctuation amplitude. Under the rolling condition with short period, the fluctuation amplitudes of the thermal parameters increase and their average values change rapidly as the rolling period decreases. Under the rolling condition with large period, the fluctuation amplitudes of the thermal parameters increase with the rolling period, and they tend to fixed values.

Experimental Research on Steam Jetting and Condensation in Low Sub-Cooled Water

2018 ◽  
Author(s):  
Chunhui Dai ◽  
Mengran Liao ◽  
Qi Xiao ◽  
Jun Wu ◽  
Shaodan Li ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Mass Flow Rate ◽  
Mass Flow ◽  
Flow Rate ◽  
Nuclear Power ◽  
Great Influence ◽  
Low Frequency ◽  
Two Phase ◽  
Noise Characteristics

Steam submerged jetting is an important process in depressurization tank and condenser deaerator tank of nuclear power plant. As the steam contact the liquid water directly, some complicated behaviors such as strong turbulence and phase transition would happens. Especially when the sub-cooling degree is low, the condensation may cause vigorous pressure pulsation and radiation noise, which not only causes noise damage to workers but also affect the safety of the heat exchanger tubes bundle because of vibration transmission. An experiment is proposed to study the complex evolutionary behavior and vibration and noise characteristics of gas-water two-phase flow. The experimental results show that in the case of low subcooling, the mass flow rate of steam has a great influence on gas plume, and, as the mass flow rate increases, the main contribution frequency of noise is gradually increasing from low frequency to high frequency. The researches in this paper can provide the technical basis for the design of the deoxygenation system of condenser in onshore and ship nuclear power plant.

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