scholarly journals Phenomenon of Coolant Local Natural Circulation Occurring in Heat Removal Loops of Nuclear Power Plant

2011 ◽  
Vol 48 (4) ◽  
pp. 602-611 ◽  
Author(s):  
Yury ASHURKO ◽  
Gennady PUGACHEV
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Heat Removal

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.

scholarly journals Seismic safety evaluation during site selection for the nuclear power plants in Bangladesh

2018 ◽  
Vol 4 (4) ◽  
pp. 251-256 ◽  
Author(s):  
Sergey Shcheklein ◽  
Ismail Hossain ◽  
Mohammad Akbar ◽  
Vladimir Velkin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Good Time ◽  
Diagnostics System ◽  
Seismic Impacts

Bangladesh lies in a tectonically active zone. Earlier geological studies show that Bangladesh and its adjoining areas are exposed to a threat of severe earthquakes. Earthquakes may have disastrous consequences for a densely populated country. This dictates the need for a detailed analysis of the situation prior to the construction of nuclear power plant as required by the IAEA standards. This study reveals the correlation between seismic acceleration and potential damage. Procedures are presented for investigating the seismic hazard within the future NPP construction area. It has been shown that the obtained values of the earthquake’s peak ground acceleration are at the level below the design basis earthquake (DBE) level and will not lead to nuclear power plant malfunctions. For the most severe among the recorded and closely located earthquake centers (Madhupur) the intensity of seismic impacts on the nuclear power plant site does not exceed eight points on the MSK-64 scale. The existing predictions as to the possibility of a super-earthquake with magnitude in excess of nine points on the Richter scale to take place on the territory of the country indicate the necessity to develop an additional efficient seismic diagnostics system and to switch nuclear power plants in good time to passive heat removal mode as stipulated by the WWER 3+ design. A conclusion is made that accounting for the predicted seismic impacts in excess of the historically recorded levels should be achieved by the establishment of an additional efficient seismic diagnostics system and by timely switching the nuclear power plants to passive heat removal mode with reliable isolation of the reactor core and spent nuclear fuel pools.

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.

Level 2 PSA Overview of HPR1000 Nuclear Power Plant

2020 ◽  
Author(s):  
Wang Ziguan ◽  
Lu Fang ◽  
Yang Benlin ◽  
Chen Shi ◽  
Hu Lingsheng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Heat Removal ◽  
Accident Prevention ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Injection System ◽  
Risk Level ◽  
Level 2

Abstract Risk-informed design approaches are comprehensively implemented in the design and verification process of HPR1000 nuclear power plant. Particularly, Level 2 PSA is applied in the optimization of severe accident prevention and mitigation measures to avoid the extravagant redundancy of system configurations. HPR1000 preliminary level 2 PSA practices consider internal events of the reactor in the context of at-power condition. Severe accidents mitigation and prevention system and its impact on the overall large release frequency (LRF) level are evaluated. The results showed that severe accident prevention and mitigation systems, such as fast depressurization system, the cavity injection system and the passive containment heat removal system perform well in reducing LRF and overall risk level of HPR1000 NPP. Bypass events, reactor rapture events, and the containment bottom melt-through induced by MCCI are among the dominant factors of the LRF. The level 2 PSA analysis results indicate that HPR1000 design is reliable with no major weaknesses.

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.

CFD Analysis of Thermal Stratification Phenomena in the RNS Suction Line of a Gen-III Nuclear Power Plant

2013 ◽  
Author(s):  
Yan Li ◽  
Daogang Lu ◽  
Zhigang Wang ◽  
Jian Wu ◽  
Fengyun Yu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Heat Removal ◽  
Radial Temperature ◽  
Piping Systems ◽  
Suction Line ◽  
Power Operation

Thermal stratification phenomena in piping systems of nuclear power plant would threaten the structural integrity of pipes, which are caused by the significant change of water density with temperature. To provide temperature gradients for the stress analysis of Normal heat Removal System (RNS) suction line of a Gen-III nuclear power plant, the relevant thermal stratification phenomena are analyzed by CFD in this paper. Cases without leakage (normal power operation) and with leakage are both studied. The results show that the first portion of pipe (one meter or so) near the hot leg is isothermal for normal power operation due to the penetrating flow. In the remaining portion, the radial temperature drops are of the order of 20∼27 K for no leakage case. For the leakage case, the radial temperature drops are 23 K or less, which are relatively smaller than those for the no leakage case due to the net hot flow from the hot leg to the valve.

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.

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