scholarly journals Study of Parametric Interactions in the Nuclear Reactor Control with Feedback

2021 ◽  
Vol 64 (6) ◽  
pp. 517-524
Author(s):  
V. V. Opiatiuk ◽  
I. L. Kozlov ◽  
V. I. Skalozubov ◽  
I. A. Ostapenko
Keyword(s):  
Relaxation Time ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Sink ◽  
Nuclear Power ◽  
Power Reactor ◽  
Operating Conditions ◽  
External Influence ◽  
Nuclear Power Reactor ◽  
Operating Modes

This article considers the principal theoretical possibility of regulating a nuclear power reactor under changing operating modes conditions when external periodic disturbances take place in conditions of changing the operating mode. By the external periodic perturbation a downward change in the conditions of the heat sink was meant. The magnitude of the changes was preliminarily calculated in such a way that the operating conditions of the power plant did not exceed the boundaries of the safe operation zone of the reactor. In the case of approaching the operation parameters to the critical ones, the heat sink was increased until the working conditions returned to their previous state. In this work the amplitude frequency response of a non-linearly enhanced system in the nuclear power plant operating conditions when non-linearly reacting to external periodic influences has been studied. The external cyclic disturbances effect produced on the reactor that initially existed under stationary operating conditions has been considered. The research was carried out by numerical simulation of the competition between processes occurring in a nuclear power plant and determined by the system’s reaction time and relaxation time while responding to periodic external influences. Calculations of the relaxation time dependence on the fixed frequency-revealing external influence’s temperature are presented. Also, the relaxation time dependence on the frequency of external influence at a fixed temperature for systems with various relaxation periods was calculated. It is determined that when the dependence between system temperature and the external influence time is calculated, there exists a wide range of possible frequency control. To evaluate the behavior of a nuclear power reactor under conditions of operating modes changes, a fundamental physical mathematical model of the reactor’s state under external harmonic influence is presented. It is based on the nonlinear Riccati equation. The external harmonic effect was simulated by changing the heat supply and heat removal conditions near the critical point.

Study on 3-D Flow Field of Concrete Volute Conduit of Circulating Water Pump in Nuclear Power Plant

2011 ◽  
Vol 99-100 ◽  
pp. 350-353
Author(s):  
Xiao Bing Sun ◽  
Xu Bin Qiao
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Stokes Equations ◽  
Three Dimensional ◽  
Gradual Change ◽  
Water Pump ◽  
Pump System ◽  
Circulating Water ◽  
Operating Modes

As the largest unit capacity of nuclear power plant at present, the flow conduit of circulating water pump in EPR1750 nuclear power plant is a volute conduit, which is a cast-in-situ conceret structure with complexly gradual change cavity. Therefore, the hydraulic efficiency of circulating water pump is not only related with the design of pump leaves, but also closely related to the design of volute and the complicated spatial type of intake and outtake conduits. With the pump leaves and the intake and outtake conduits of conceret volute as the research model, based on computational fluid dynamics (CFD)and the three dimensional Reynolds averaged Navier-Stokes equations, an analytic model suitable for computation is established to simulate the three-dimensional steady flow in the whole pumping system under different operating modes. By use of the commercial fluid-computation softer ANSYS, the distribution of basic physic quantities in the fluid field inside the pump and the conduits is obtained. The analysis and prediction of the performance of pump system are made, and the spatial type design of intake and outtake conduits is evaluated. The calculation results can be referenced to improve the design of pump systems in the similar projects.

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.

Ultimate Heat Sink Sizing Requirements on a Month-to-Month Basis

2009 ◽  
Author(s):  
Dong Zheng ◽  
Allen T. Vieira ◽  
Julie M. Jarvis ◽  
George P. Emsurak
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Sink ◽  
Nuclear Power ◽  
Meteorological Data ◽  
Meteorological Condition ◽  
Water Requirement ◽  
Next Generation ◽  
Worst Case ◽  
Automated Methods

The Ultimate Heat Sink (UHS) of a nuclear power plant is a complex cooling water system which serves the plant during normal and accident conditions. For some next generation nuclear plants, the UHS sizing is a major design and licensing analysis task. The analysis involves detailed modeling of the transient heat loads and the selection of worst-case meteorological data for the plant site. The UHS sizing requirements for a representative next generation nuclear power plant are evaluated on a month-to-month basis. This paper assesses the UHS water requirement for each month of year. The UHS analysis for a representative next generation nuclear plant with mechanical draft cooling towers and a water basin is used to determine the maximum evaporation of the basin for the worst-case meteorological data on a month-to-month basis. To size the cooling tower basin, automated methods have been developed which determine the highest evaporative losses from the basin and highest basin temperature over a 30-day design basis accident period. This paper also evaluates the month-to-month basin temperature changes. This assessment is done for a representative next generation nuclear power plant and considers the monthly historical meteorological data over 45 years. The result of this assessment of monthly UHS water requirement is of interest in assessing the margin in the UHS design. This monthly assessment is also useful in demonstrating that the automated methods used to establish the limiting 30-day meteorological condition are indeed accurate. In addition, these results may be useful in helping to plan plant maintenance activities.

Centrifugal Pump Leak-Free Technology Introduction and Its Applications in EPR Nuclear Power Plant

2010 ◽  
Author(s):  
Guohui Cong ◽  
Ling Zhang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Centrifugal Pump ◽  
Nuclear Power ◽  
Operating Conditions ◽  
Critical Conditions ◽  
Mechanical Seal ◽  
Centrifugal Pumps ◽  
Canned Motor ◽  
Magnetic Drive

Environmental protection requirement is more and more critical now, and it increases the request to prevent dangerous liquid to leak outside in nuclear power plant too. Centrifugal pumps are the most important active equipments in nuclear power plant, but there is a shaft clearance between rotor and stator of centrifugal pump. The shaft clearance can lead pumped fluid to the outside, so the environment may be polluted by the leakage. In some critical conditions such as transferring high radioactive fluid in the pump, the leakage shall be totally forbidden. So solutions have to be found to make centrifugal pumps totally leak-free for applications in nuclear power plant. Normally there are three leak-free technologies for centrifugal pumps: mechanical seal with auxiliary system, canned motor and magnetic drive. In this paper, all the three leak-free technologies and some of their applications in EPR 3rd generation PWR nuclear power plants are presented and discussed. The results show that in EPR nuclear power plant, canned motor pumps can be preferably used for strict environmental requirement of leak-free if the pump power and operating conditions are applicable. For other conditions, pumps with double mechanical seal can also be used with additional sealing water system support. For centrifugal pumps with magnetic drive are not so applicable in high pressure condition, and the safety aspect is weaker than canned motor pumps, generally they are not used in EPR nuclear power plant at present.

Research on Operating Strategy of Total Loss of Essential Service Water System With RHR Not Connected in PWR Unit

2020 ◽  
Author(s):  
Shengtao Zhang ◽  
Ke Yi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Sink ◽  
Nuclear Power ◽  
Cooling System ◽  
Water System ◽  
Total Loss ◽  
Operating Strategy ◽  
Service Water ◽  
Essential Service

Abstract Essential Service Water System (WES) is part of the nuclear power plant cooling system which provides the final heat sink for nuclear power plants. Therefore, WES must operate stably, safely and reliably for a long time. The total loss of WES accident is a design extended condition and will result in the loss of the final heat sink of the unit. The consequences of the accident are severe. In order to deal with the accident quickly and effectively and ensure the safety and economics of the power plant in accident condition, it’s necessary to formulate corresponding treatment strategy to deal with the transient. This paper developed a strategy for dealing with the total loss of WES with Residual Heat Removal System (RHR) not connected condition in Generation III nuclear power plant. The structure of the WES system and the types of failures that may occur are analyzed, and thus the symptoms of the faults are obtained and the entry conditions for the operating strategy are determined. The effect of faults on unit equipment and safety functions and the impact on nuclear steam supply system (NSSS) control are analyzed in this paper. Combined with the unit design, the system and equipment for controlling and mitigating related safety functions are analyzed, and the mitigation method and the fallback strategy of the fault are determined. Thereby a complete operating strategy of total loss of WES with RHR not connected is obtained. In addition, this paper analyzes and evaluates the operating strategy by simulating thermal hydraulic calculation for the first time. The results show that without staff intervention Component Cooling System (WCC) temperature reached 55°C limits after running a few minutes. Based on the intervention of the operating strategy proposed in this paper, WCC temperature reached the 55°C limits when the unit was operated at about 4 hours and 55 minutes. The result shows that and the strategy can effectively alleviate the failure and provide sufficient intervention time for the operator to bring the unit to a safe state.

Study on the Operating Strategy Optimization of Moving to Remote Shutdown Station When Main Control Room Is Un-Inhabitable

2018 ◽  
Author(s):  
Li Li ◽  
Zhang Shengtao ◽  
Xu Zhao ◽  
Du Yu
Keyword(s):  
Remote Control ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Operating Conditions ◽  
Operation Strategy ◽  
Control Room ◽  
Forced Circulation ◽  
Operating Strategy ◽  
Power Operation

For PWR, remote shutdown station (RSS) is a redundant control mean to shut down the reactor when main control room (MCR) inhabitation is challenged (e.g. fire, smoke...). Nowadays, due to nuclear power plants control measures were improved with DCS system, a full function DCS RSS was equipped and more essential equipment could be controlled on RSS. Under operating conditions that prohibit nuclear power plant operators to stay in the main control room, the operators should move to RSS and shutdown the reactor to ensure plant safety following <Moving to remote shutdown station when main control room is un-inhabitable operating strategy> (RSS strategy for short) to fallback the plant from power operation to cold shutdown. The original operating strategy by nature circulation is no longer the best choice both for operation safety and economy efficiency, and an optimized new strategy should be raised. Based on the former reason, an optimized operation strategy was raised in this paper. In the optimized strategy, all plant normal standard operation modes were considered as initial conditions, rather than only considering power operation condition in the original one. The fallback mode and fallback strategy for each initial condition was also designed and optimized. To accelerate the depression and heat removal process, a forced circulation operation strategy is adopted when the reactor coolant pumps are available, and less local operation was included by taking advantages of the full function operating measures on RSS. To simplify the whole procedure structure, the operation modules of other general operating procedures are reused. To validate the effectiveness of the optimized operating strategy, a full scope PWR simulation tool was employed to make thermo hydraulic calculation validation of the reactor response and also the remote control station HMI supporting validation. By simulating the original strategy and the optimized one and related analysis, we found that the optimized strategy is effective, and able to be executed based on the remote control station hardware. By executing the optimized strategy, the unit can fall back to the cold shutdown condition safely and a few hours were saved compared with the original strategy. The optimized strategy had already been implemented on real PWR nuclear power plant.

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.

Performance of a Spray Pond for Nuclear Power Plant Ultimate Heat Sink

1977 ◽  
Vol 99 (4) ◽  
pp. 650-656
Author(s):  
V. E. Schrock ◽  
G. J. Trezek ◽  
L. R. Keilman
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Performance ◽  
Heat Sink ◽  
Nuclear Power ◽  
Field Tests ◽  
Computer Code ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Transient Thermal

Spray ponds have become an attractive method of providing the “ultimate heat sink”, i.e., the assured means of dissipating heat from a nuclear power plant. Two redundant spray ponds were the choice for this service in the Rancho Seco Nuclear Generating Station owned by Sacramento Municipal Utility District. This paper describes the results of full scale field tests of the Rancho Seco ponds which were conducted to verify the thermal performance, drift loss characteristics, and the capability to sustain the cooling requirements for a period of 30 days following a loss-of-coolant accident (LOCA). Correlations of local and average nozzle efficiency and of the drift loss are presented. A computer code was developed for the transient thermal performance of the pond. After verification the code was used to predict performance following LOCA under adverse meteorological conditions based on weather records.

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