scholarly journals Analysis on Steady-State Operation and Heat Loss of Chinese Integrated Pressurized Water Reactor

2014 ◽  
Vol 2014 ◽  
pp. 1-7
Author(s):  
Zhang Fan ◽  
Lu Dao-Gang ◽  
Sui Dan-Ting ◽  
Guo Chao ◽  
Yuan Bo
Keyword(s):  
Steady State ◽  
Heat Loss ◽  
Heat Balance ◽  
Power Plants ◽  
Thermal Power ◽  
Balance Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Steady State Operation

Chinese integrated pressurized water reactor (CIPWR) has compact configuration and high inherent safety, which is appropriate for nuclear power plants of small and medium scale. Heat balance model has been adopted widely in thermal power calibration of PWRs because of its advantage of accuracy. In this paper, a package based on FORTRAN language is developed and added into RELAP5 to calculate the heat loss value needed in heat balance analysis. The steady-state operation of CIPWR is modelled correctly by RELAP5. The heat loss of CIPWR is calculated by the package, and the comparison of the main values of parameters needed in the heat loss calculation between RELAP5 and the package has been done. It shows that the package has high calculation accuracy and can be applied in reactor design and monitoring.

Radionuclide Composition of Airborne Discharges of European NPPS With WWER, Pressurized Water Reactor, and Boiling Water Reactor Facilities

2020 ◽  
Vol 7 (1) ◽  
Author(s):  
Angelina-Nataliya V. Vukolova ◽  
Andrei A. Rusinkevich
Keyword(s):  
Power Plants ◽  
Boiling Water ◽  
Normal Operation ◽  
Boiling Water Reactor ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Operation Conditions ◽  
Discharge Activity ◽  
Radionuclide Composition

Abstract The article presents the analysis of the data on radionuclide composition of airborne discharges of 52 European nuclear power plants (NPPs) with water–water energetic reactor facilities (WWER), pressurized water reactor facilities (PWR), and boiling water reactor facilities (BWR) under normal operation conditions. It contains lists of radionuclides, registered in discharges of researched NPPs, and gives estimation of contributions of radionuclides, forming the discharge, into total activity of discharge and into total effective dose, created by the discharge activity. It was determined that the maximal contribution into discharge activity of all researched NPPs make noble gases, tritium, and carbon-14, while the latter is the main dose-making radionuclide.

Stress Corrosion Cracking Behavior of Alloy 182 Weld in Pressurized Water Reactor Primary Water Environment at 325ºC

CORROSION ◽  
2011 ◽  
Vol 67 (8) ◽  
pp. 085004-1-085004-9 ◽  
Author(s):  
L.I.L. Lima ◽  
M.M.A.M. Schvartzman ◽  
C.A. Figueiredo ◽  
A.Q. Bracarense
Keyword(s):  
Stress Corrosion ◽  
Oxide Layer ◽  
Stress Corrosion Cracking ◽  
Power Plants ◽  
Corrosion Cracking ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Cracking Behavior ◽  
Water Reactor ◽  
Primary Water

Abstract The weld used to connect two different metals is known as a dissimilar metal weld (DMW). In nuclear power plants, this weld is used to join stainless steel to low-alloy steel components in the nuclear pressurized water reactor (PWR). The most common weld metal is Alloy 182 (UNS W86182). Originally selected for its high corrosion resistance, it exhibited, after a long operation period, susceptibility to stress corrosion cracking (SCC) in PWR. The goal of this work was to study the electrochemical corrosion behavior and SCC susceptibility of Alloy 182 weld in PWR primary water containing 25 cm3 and 50 cm3 H2/kg H2O at standard temperature and pressure (STP). For this purpose, slow strain rate tensile (SSRT) tests and potentiodynamic polarization measurements were carried out. Scanning electron microscopy (SEM) with energy-dispersive spectrometry (EDS) was used to evaluate fracture morphology and determine the oxide layer chemical composition and morphology. The results indicated that at 325°C Alloy 182 weld is more susceptible to SCC at 25 cm3 (STP) H2/kg H2O and the increase of dissolved hydrogen decreased the crystal size of the oxide layer.

Stability of Cement-Glass Packages Containing Sodium Borate Salt Generated from Pressurized Water Reactor Power Plants

1987 ◽  
Vol 78 (2) ◽  
pp. 185-190 ◽  
Author(s):  
Tatsuo Izumida ◽  
Fumio Kawamura ◽  
Koichi Chino ◽  
Makoto Kikuchi
Keyword(s):  
Power Plants ◽  
Reactor Power ◽  
Sodium Borate ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

Small Modular Pressurized Water Reactors Combined With Conventional Thermal Power Plant

2017 ◽  
Author(s):  
Huijing Jiang ◽  
Ning Bai ◽  
Huanfen Zhan ◽  
Feng Shen ◽  
Bin Gao ◽  
...  
Keyword(s):  
Power Plant ◽  
Thermal Power Plant ◽  
Nuclear Energy ◽  
Consumption Rate ◽  
Thermal Power ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Hybrid Power ◽  
Hybrid Power Plant

Small modular reactor is investigated worldwide with the advantages of lower initial investment and short construction period. Generally, the economy of small modular pressurized water reactor (PWR) is not as good as large PWR, so various applications of small PWR are investigated, such as marine reactor, heat supply and sea water desalination. Limited to the parameters of steam generator, the generating efficiency for the pressurized water reactor nuclear power plant is about 33%, while the steam temperature of supercritical fossil power plant can exceed 600°C and generating efficiency is more than 45%. The essence of a hybrid power plant is to use a fossil fuel to superheat wet steam in an outer steam superheating device, after the steam generator to improve the parameters of working fluid. On one hand, the innovative hybrid nuclear power plant which combines nuclear reactor with conventional thermal energy can improve the efficiency of small PWR. On the other hand, this hybrid power plant has lower carbon emission compared with traditional thermal power plant. This paper describes two different coupling schemes of small pressurized water reactor combined with supercritical thermal power plant using steam turbine. Efficiency of hybrid power plant is influenced by the coupling scheme, steam parameter of the superheating device outlet, the proportion of nuclear energy, efficiency of assemblies and so on. The plant efficiency becomes higher with the improvement of parameter of the superheating device’s outlet steam, and it is higher when the proportion of nuclear energy becomes lower. Take the 660MWt integrated small pressurized water reactor as an example, when the proportion of nuclear energy accounts for 48%, the thermal efficiency of this innovative hybrid power plant is about 43%,while the net efficiency is 41%, that is much higher than the efficiency of traditional pressurized water reactor, improving about 24 percent. As to the carbon emission, it depends on the coal consumption rate of power supply. The coal consumption rate of this hybrid power plant is 158g /kWh, while the consumption rate of thermal hybrid power plant is 280g/kWh, reducing about 44 percent. Also, the fundamental solutions of technical problems for this innovative hybrid power plant are discussed in the paper. Furthermore, several useful outcomes and suggestions for key equipments are put forward, such as the scheme of a superheating device and high-temperature steam turbine, and the possibility of using a lava boiler as the superheating device to improve the steam parameters after the steam generator is analyzed.

scholarly journals Analysis of thermal hydraulic behavior of KONVOI PWR during a design extension condition

2021 ◽  
pp. 4-4
Author(s):  
Salwa Helmy ◽  
Magy Kandil ◽  
Ahmed Refaey
Keyword(s):  
Power Plants ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Hydraulic Behaviour ◽  
Loss Of Coolant ◽  
Safety Systems

In Nuclear Power Plants the Design Extension Conditions are more complex and severe than those postulated as Design Basis Accidents, therefore, they must be taken into account in the safety analyses. In this study, many hypothetical investigated transients are applied on KONVOI pressurized water reactor during a 6-in. (182 cm2) cold leg Small Break Loss-of-Coolant-Accident to revise the effects of all safety systems ways through their availability/ nonavailability on the thermal hydraulic behaviour of the reactor. The investigated transients are represented through three cases of Small Break Loss-of-Coolant-Accident as, case-1, without scram and all of the safety systems are failure, case-2, the normal scram actuation with failure of all safety systems (nonavailability), and finally case 3, with normal actuation scram sequence and normal sequential actuation of all safety systems (availability). These three investigated transient cases are simulated by creation a model using Analysis of Thermal-Hydraulics of LEaks and Transient code. In all transient cases, all types of reactivity feedbacks, boron, moderator density, moderator temperature and fuel temperature are considered. The steady-state results are nearly in agreement with the plant parameters available in previous literatures. The results show the importance effects of the feedbacks reactivity at Loss-of-Coolant-Accident on the fallouts power, since they are considered the key parameters for controlling the clad and fuel temperatures to maintain them below their melting point. Moreover, the calculated results in all cases show that the thermal hydraulic parameters are in acceptable ranges and encounter the safety criterion during Loss-of-Coolant-Accident the Design Extension Conditions accidents processes. Furthermore, the results show that the core uncovers and fuel heat up do not occur in KONVOI pressurized water reactor in theses the Design Extension Conditions simulations since, all safety systems provide adequate core cooling by sufficient water inventory into the core to cover it.

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