The core shroud leakage analysis and study for Kuosheng nuclear power plant

Kerntechnik ◽  
2019 ◽  
Vol 84 (3) ◽  
pp. 161-168
Author(s):  
H.-T. Lin ◽  
J.-R. Wang ◽  
H.-C. Chen ◽  
J.-H. Yang ◽  
S.-W. Chen ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
The Core ◽  
Analysis And Study ◽  
Core Shroud

The Three Mile Island (TMI) Nuclear Accident: Revisited

1985 ◽  
Vol 1 (S1) ◽  
pp. 401-404
Author(s):  
Donald Reid
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Test Tube ◽  
Human Errors ◽  
The Core ◽  
Core Damage ◽  
Short Period ◽  
Water Pumps ◽  
The Universe

At 0400 hours on Wednesday, March 28, 1979, an extremely small and initially thought unimportant malfunction occurred at the nuclear power plant at Three Mile Island (TMI). Within a short period of time, that malfunction would turn into an event of momentous impact with repercussions felt over most of the world. The events of that malfunction would cause TMI to be labelled as the worst commercial nuclear incident in history and transform it into the nuclear test tube of the universe. What really happened at Three Mile Island? Thirty-six seconds after 0400 hours, several water pumps stopped functioning in the unit 2 nuclear power plant. In the minutes, hours and days that followed, a series of events—compounded by equipment failure, inappropriate procedures and human errors—escalated into the worst crisis yet experienced by the nation's nuclear power industry. This resulted in the loss of reactor coolant, overheating of the core, damage to the fuel (but probably no melting) and release outside the plant of radioactive gases. Hydrogen has was formed, primarily by the reaction between the zirconium casing that holds the radioactive fuel and steam. There, however, was no danger of the bubble inside the reactor vessel exploding, because of the absence of oxygen within the reactor.

scholarly journals Determination of the safety rods (SA, SB) for optimized power reactor 1000 using the Core simulator OPR1000

2018 ◽  
Vol 1 (6) ◽  
pp. 177-184
Author(s):  
Son An Nguyen ◽  
Nguyen Trung Tran
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Boron Concentration ◽  
Power Reactor ◽  
Experimental Simulation ◽  
The Core ◽  
Operation Process

In order to operate a nuclear power plant, ensuring safety is the most important factor. The function of safety rods are to shut down the reactor in case of emergency. The purpose of this paper to show the result of research and determine the value of safety rods SA, SB. Determination of the Boron concentration corresponding to each group of safety rods of OPR1000 nuclear reactor ensures the safely in the whole operation process. Experimental simulation is carried out in the system simulating core reactor OP1R1000 (CoSi OPR1000). The expermental result corresponds with the theoretic calculated result of Sa and Sb with 1500 pcm, 4000 pcm. The concentrations of Boron appropriately are 134 ppm and 284 ppm, respectively.

Deposition Fractions of Fission Product and Heavy Elements on Soil in Fukushima Dai-Ichi Nuclear Power Plant Accident

2013 ◽  
Author(s):  
Toru Yamamoto
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Soil Samples ◽  
Heavy Elements ◽  
Severe Accident ◽  
Accident Analysis ◽  
Radioactivity Measurement ◽  
Power Station ◽  
The Core

Based on radioactivity measurement of soil samples in the site of Fukushima Dai-Ichi Nuclear Power Station, radioactivity of Sr, Nb, Mo, Tc, Ru, Ag, Te, I, Cs, Ba, La, Pu, Am, and Cm isotopes were compiled as radioactivity ratios to 137Cs. By exponentially fitting or averaging, the radioactivity ratios at the core shutdown were estimated. They were divided by those of the fuel of the core at the shutdown to obtain a deposited radioactivity fractions of the nuclides as relative values to 137Cs, which also correspond to deposition fractions of the elements as relative values to Cs. They were estimated to be orders of 10−4 to 10−3 for Sr, 10−4 for Nb, 10−2 to 10−1 for Mo, 10−1 for Ag, 10−1 to 100 for Te, 100 for I, 10−3 for Ba, 10−6 to 10−5 for Pu, 10−6 to 10−5 for Am, and 10−6 for Cm. The observed radioactivity ratios to 137Cs were compared with those obtained by severe accident analysis to assess the validation of the analysis.

1/3D modeling of the core coolant circuit of a PHWR nuclear power plant

2015 ◽  
Vol 83 ◽  
pp. 386-397 ◽  
Author(s):  
Santiago Corzo ◽  
Damian Ramajo ◽  
Norberto Nigro
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
3D Modeling ◽  
Nuclear Power ◽  
Coolant Circuit ◽  
The Core

The Analysis of AP1000 Beyond Design Basis DEDVI Accident

2013 ◽  
Author(s):  
Sheng Zhu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Heat Removal ◽  
The Core ◽  
Design Basis ◽  
Injection Line ◽  
Design Basis Accident ◽  
Core Cooling

Double ended break of direct vessel injection line (DEDVI) is the most typical small-break lost of coolant accident (LOCA) in AP 1000 nuclear power plant. This study simulated the DEDVI (without actuation of automatic depressurization system 1–3 stage valves, accumulators and passive residual heat removal heat exchanger) beyond design basis accident (BDBA) to validate the safety capability of AP1000 under such conditions. The results show that the core will be uncovered for about 863 seconds and then recovered by water after gravity injection from IRWST into the pressure vessel. The peak cladding temperature (PCT) goes up to 838.08°C, much lower than the limiting value 1204°C. This study confirms that in the DEDVI beyond design basis accident, the passive core cooling system (PXS) can effectually cool the core and preserve it integrate, and ensure the safety of AP 1000 nuclear power plant.

Research and Engineering Application of Repaired Fuel Assembly Reusing Technique

2017 ◽  
Author(s):  
Lihua Wang ◽  
Qingxiang Yang ◽  
Ping Yang ◽  
Jiazheng Liu ◽  
Libing Zhu ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Engineering Application ◽  
Evaluation Technique ◽  
Fuel Rods ◽  
The Core ◽  
Fuel Assemblies

Due to debris in the coolant against clad, fuel clad wear, fuel handling fault and so on, fuel rods maybe be damaged during the operation of nuclear power plants, in order that the fuel assemblies with damaged fuel rods are discharged before scheduled. If the damaged fuel assemblies are not reloaded into the core of the nuclear power plant, the fuel utilization decreases and the economy of the nuclear power plant is partly lost. For retrieving the loss of the economy, the damaged fuel assemblies can be repaired by replacing damaged fuel rods with dummy rods which don’t include fissile nuclides. Then, the repaired fuel assemblies can be reloaded into the core. As the repaired fuel assemblies are different with the normal fuel assemblies, especially the number of the damaged fuel rods is considerable, a whole quantitative analysis is very necessary to evaluate the effects from the reuse of the repaired fuel assemblies. In this paper, a full scope evaluation of reload design are performed including nuclear design, fuel design, thermal hydraulic design and safety evaluation, and some necessary improvements are done for the software system, design methods and progress which have been used in the normal reload design. As results, an integrated evaluation technique is developed to evaluate the feasibility and safety of reusing the repaired fuel assemblies, and the key effects due to the reuse of the repaired fuel assemblies are extracted, and the different effects are studied for the different materials of the dummy rods which can be used to conduct how to choose the proper material of dummy rods. In addition, this technique has been successfully applied in the engineering and the loss of economy due to the damage of fuel assemblies was retrieved partly. Therefore, the integrated evaluation technique has also important directive to other nuclear power plants if the repaired fuel assemblies are planned to reuse.

Melting Simulation Using a Particle Method With Angular Momentum Conservation

2017 ◽  
Author(s):  
Masahiro Kondo ◽  
Shota Ueda ◽  
Koji Okamoto
Keyword(s):  
Angular Momentum ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Particle Method ◽  
Momentum Conservation ◽  
Severe Accident ◽  
Angular Momentum Conservation ◽  
The Core ◽  
Melting And Solidification

To analyze the core degradation and relocation behavior of melts in a severe accident of nuclear power plant, the melting and solidification in the complexed geometry is to be calculated. For the calculation of such complexed behavior, a new particle method conserving angular momentum is proposed and applied for the melting simulation. When solid melts, it may move like a rigid body. The angular momentum conservation is important to capture such kind of motion. The potential of the new particle method was confirmed with a calculation of the melting in dam break geometry and cantilever geometry.

Uncertainty Analysis of SB-LOCA in AP1000 Nuclear Power Plant

2017 ◽  
Author(s):  
Hao Shi ◽  
Qi Cai ◽  
Yuqing Chen ◽  
Lizhi Jiang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Hydraulic System ◽  
Nuclear Power ◽  
Safety Margin ◽  
Analysis Model ◽  
The Core ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Best estimate plus uncertainty methods (BEPUs) could be used to eliminate the over-conservatism and gain more safety margin in the analysis of thermal-hydraulic transient process at nuclear power plant. Based on the Best estimate thermal-hydraulic system code RELAP5/MOD3.2 platform, the best estimate plus uncertainty methods (BEPUs) proposed by GRS (Gesellschaft fur Anlagen- and Reaktorsicherheit) are presented together with applications to a small break loss of coolant accident (SB-LOCA) on the AP1000 Nuclear Power Plant best estimate analysis model. According to the results of uncertainty calculations, the dispersion bands of maximum cladding temperature and the core outlet void fraction are displayed and assessed.

Study on Shutdown Fire PRA for Nuclear Power Plant

2018 ◽  
Author(s):  
Meiru Liu ◽  
Qingnan Zhao ◽  
Wei Deng ◽  
Jinyan Du ◽  
Lin Sun
Keyword(s):  
Risk Assessment ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
East Coast ◽  
Fire Risk ◽  
The Core ◽  
Core Damage ◽  
Risk Result

Fire Probabilistic Risk Assessment (PRA) is one of the main methods of fire safety analysis for nuclear power plants (NPPs). At present, the fire PRA under the at-power condition has been widely studied, while the research on the low power and shutdown condition (LPSD) is quite limited. Therefore, in this paper, a second generation NPP on the east coast of China is taken as the research target, and the analysis methods are based on the latest LPSD fire PRA theory in report NUREG/CR-7114. This paper studies the initiating events and ignition frequencies of fire PRA considering the real conditions in LPSD, and established LPSD Fire PRA model, finally obtained the quantitative risk result of the core damage caused by the fire According to the results of this LPSD fire PRA, the fire risk-significant sources and fire risk weakness are found out and the improvement suggestions have been promoted.

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