Controlling and Supervising the Nuclear (Core) Power of a PWR in Relation to Nuclear Safety and Economics

2009 ◽  
Author(s):  
P. Wouters ◽  
W. Van Rompay ◽  
F. Bertels ◽  
W. Van Hove ◽  
E. Gorleer ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Total Error ◽  
Payback Period ◽  
Measuring System ◽  
Steam Generators ◽  
Nuclear Core ◽  
Ultrasonic Flow Meter

Knowing exactly the nuclear core power of a nuclear reactor is one of the most important parameters for the operator; it is vital for safety as well as for economical matters. The secondary calorimetric is the only one where one can pilot on; it is a combination of measured parameters, of which the feedwater (FW) flow towards the steam generators is the most significant one. This feedwater flow can be measured by means of an ultrasonic flow meter, “LEFM CheckPlus™ system” instead of the commonly used venturis or diaphragms. In the Belgian Nuclear Power Plant (NPP) Doel 4, a new ultrasonic “LEFM CheckPlus™” feedwater flow measuring system has been installed in April 2008. The paper describes the consequences of the installation, as the total error on the secondary calorimetric decreases from the previous 1,3% to the current 0,8% with a possibility of further reduction to 0,4%. Additionally, the economical effects of the installation are calculated for a 1000 MWe power plant with venturi meters undergoing fouling. For the NPP Doel 4 it was an economically interesting investment since the payback period was only 45 days. Finally, the possibility of consuming the margin on the secondary calorimetric for a mini-power uprate is inspected, technically and economically. It is concluded that such a mini-power uprate is an interesting option for the NPP owner.

A High Effective Parallel Method for the Coupling Between Neutronics and Thermal-Hydraulic

2016 ◽  
Author(s):  
Xiaomeng Dong ◽  
Zhijian Zhang ◽  
Zhaofei Tian ◽  
Lei Li ◽  
Guangliang Chen
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Decisive Role ◽  
Reactor Power ◽  
Coupling Method ◽  
Outlet Temperature

Multi-physics coupling analysis is one of the most important fields among the analysis of nuclear power plant. The basis of multi-physics coupling is the coupling between neutronics and thermal-hydraulic because it plays a decisive role in the computation of reactor power, outlet temperature of the reactor core and pressure of vessel, which determines the economy and security of the nuclear power plant. This paper develops a coupling method which uses OPENFOAM and the REMARK code. OPENFOAM is a 3-dimension CFD open-source code for thermal-hydraulic, and the REMARK code (produced by GSE Systems) is a real-time simulation multi-group core model for neutronics while it solves diffusion equations. Additionally, a coupled computation using these two codes is new and has not been done. The method is tested and verified using data of the QINSHAN Phase II typical nuclear reactor which will have 16 × 121 elements. The coupled code has been modified to adapt unlimited CPUs after parallelization. With the further development and additional testing, this coupling method has the potential to extend to a more large-scale and accurate computation.

scholarly journals Neutronic safety analysis of proposed reactor technologies for Ghana’s nuclear power plant using the MCNP code

2019 ◽  
Vol 34 (3) ◽  
pp. 238-242
Author(s):  
Rex Abrefah ◽  
Prince Atsu ◽  
Robert Sogbadji
Keyword(s):  
High Pressure ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Atomic Energy Commission ◽  
Clean Energy ◽  
Safety Analysis ◽  
Operating Conditions ◽  
The Government

In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Ghana, the Nuclear Power Institute of the Ghana Atomic Energy Commission has on the agenda to advise the government on the nuclear power to include in the country's energy mix. After consideration of several proposed nuclear reactor technologies, the Nuclear Power Institute considered a high pressure reactor or vodo-vodyanoi energetichesky reactor as the nuclear power technologies for Ghana's first nuclear power plant. As part of technology assessments, neutronic safety parameters of both reactors are investigated. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality and neutron behavior at various operating conditions. The high pressure reactor which is still under construction and theoretical safety analysis, showed good inherent safety features which are comparable to the already existing European pressurized reactor technology.

Experimental Investigation of Subcooled Choking Flow in a Steam Generator Tube Crack

2016 ◽  
Author(s):  
Hung Nguyen ◽  
Mark Brown ◽  
Shripad T. Revankar ◽  
Jovica Riznic
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Experimental Program ◽  
Steam Generators ◽  
Steam Generator Tube ◽  
Loss Of Coolant ◽  
Steam Generator Tubes

Steam generator tubes have a history of small cracks and even ruptures, which lead to a loss of coolant from the primary side to the secondary side. These tubes have an important role in reactor safety since they serve as one of the barriers between radioactive and non-radioactive materials of a nuclear power plant. A rupture then signifies the loss of the integrity of the tube itself. Therefore, choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant, but also to everyday operation. There is limited data on actual steam generators tube wall cracks. Here experiments were conducted on choked flow of subcooled water through two samples of axial cracks of steam generator tubes taken from US PWR steam generators. The purpose of the experimental program was to develop database on critical flow through actual steam generator tube cracks with subcooled liquid flow at the entrance. The knowledge of this maximum flow rate through a crack in the steam generator tubes of a pressurized water nuclear reactor will allow designers to calculate leak rates and design inventory levels accordingly while limiting losses during loss of coolant accidents. The test facility design is modular so that various steam generator tube cracks can be studied. Two sets of PWR steam generators tubes were studied whose wall thickness is 1.285 mm. Tests were carried out at stagnation pressure up to 6.89 MPa and range of subcoolings 16.2–59°C. Based on these new choking flow data, the applicability of analytical models to highlight the importance of non-equilibrium effects was examined.

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.

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.

Numerical Investigation of Aerosol Collection Efficiency in a Venturi Scrubber

2018 ◽  
Author(s):  
Paridhi Goel ◽  
Arun K. Nayak
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Collection Efficiency ◽  
Computational Framework ◽  
Venturi Scrubber ◽  
Lagrangian Framework ◽  
Air Stream ◽  
Safety Systems

In an extreme situation, as happened in Fukushima nuclear power plant, the failure of multiple safety systems may lead to severe core damage and melt relocation. This may be accompanied by production of large amount of steam which may result in the over-pressurization of the containment. Another associated concern is that the exposed core melt is highly radioactive which if exposed to the atmosphere can ruthlessly deteriorate the quality of environment and living beings. The radioactive materials present in the containment can be in vapor form or aerosol form. The containment of a nuclear power plant is therefore the final shielding to prevent the release of radioactive products to the environment. Therefore, the installation of Filtered Containment Venting system (FCVS) is mandatory in a nuclear reactor which actuates passively to depressurize the containment. Additionally it assists in the retention of radionuclides in the containment. The FCVS consists of venturi scrubbers submerged in a pool of scrubbing liquid along with a demister housed in a scrubber tank. The performance of venturi scrubber is dependent on the interaction of the contaminated air stream from the nuclear reactor with the scrubbing liquid. This represents a multi-field and multi-fluid system. The present analysis investigates this system through a computational framework in which air stream is solved using Eulerian framework while the scrubbing liquid is tracked through Lagrangian framework. The collection efficiency of aerosols is modeled assuming impaction to be the predominant mechanism.

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