scholarly journals Fatigue Strength Evaluation Of Pressurizer Wall Structure In Pressurized Water Reactor

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Roziq Himawan
Keyword(s):  
Fatigue Strength ◽  
Nuclear Power ◽  
System Analysis ◽  
Fatigue Curve ◽  
Failure Criteria ◽  
Wall Structure ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Design Qualification

<p>Fatigue strength evaluations have been performed to the pressurizer component in Pressurized Water Reactor. Fatigue is the main failure mechanism of material during system in operation. Therefore, this evaluation becomes important to be performed since the pressurizer has a very important function in the reactor’s system. Analysis was performed by using Nuclear Power Plant operation data from 40 years operation and base on Miner theory. This analysis covered all stress level experienced by the reactor during the service. To determine the value of fatigue usage factor a, fatigue curve of SA 533 material was applied. Analysis results show that the cumulative fatigue damage during 40 years in operation is 4,23×10<sup>-4</sup>. This value still far enough below failure criteria, which a value is 1. Therefore, the pressurizer design has already fulfilled the design qualification in term of fatigue aspect.</p>

scholarly journals Concentration of Radiocarbon and Its Chemical Forms in Gaseous Effluents, Environmental Air, Nuclear Waste and Primary Water of a Pressurized Water Reactor Power Plant in Hungary

Radiocarbon ◽  
1995 ◽  
Vol 37 (2) ◽  
pp. 497-504 ◽  
Author(s):  
Mihály Veres ◽  
Ede Hertelendi ◽  
György Uchrin ◽  
Eszter Csaba ◽  
István Barnabás ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Discharge Rate ◽  
Chemical Forms ◽  
Pressurized Water Reactor ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Water Reactor ◽  
Environmental Air

We measured airborne releases of 14C from the Paks Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP). Two continuous stack samplers collect 14C in 14CO2 and 14CnHm chemical forms. 14C activities were measured using two techniques; environmental air samples of lower activities were analyzed by proportional counting, stack samples were measured by liquid scintillation counting. 14C concentration of air in the stack varies between 80 and 200 Bqm−3. The average normalized yearly discharge rates for 1988–1993 were 0.74 TBqGW−1ey−1 for hydrocarbons and 0.06 TBqGW−1ey−1 for CO2. The discharge rate from Paks Nuclear Power Plant is about four times higher than the mean discharge value of a typical Western European PWR NPP. The higher 14C production may be apportioned to the higher level of nitrogen impurities in the primary coolant. Monitoring the long-term average excess from the NPP gave D14C = 3.5‰ for CO2 and D14C = 20‰ for hydrocarbons. We determined 14C activity concentration in the primary coolant to be ca. 4 kBq liter−1. The 14C activity concentrations of spent mixed bed ion exchange resins vary between 1.2 and 5.3 MBqkg−1 dry weight.

The Fault Tree Analysis on Auxiliary Feedwater Pump Room Ventilation System in Pressurized Water Reactor Power

2013 ◽  
Author(s):  
Yang Li ◽  
Chen Hang
Keyword(s):  
Nuclear Power ◽  
Fault Tree ◽  
Fault Tree Analysis ◽  
Easy Access ◽  
Main Function ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Tree Analysis ◽  
Water Reactor ◽  
Mandatory Requirement

Main function of HVAC is to remove heat from equipment and pipeline, hold the inner condition, maintain an ambient temperature and humidity that keep the equipments function properly and easy access. Although regulation is no mandatory requirement of redundant equipment design and preservation function in case of specified disaster or man-made accident. In fact, It does be influenced by the incident whether partial failure or full. The hazard factor determination and qualitative analysis are based on fault tree analysis through simulated mode from selected the typical system. The identification of accident cause, hazard cause and fault mode is essential for improving system reliability. According the analysis result, It will be optimization factor such as installation and design process, maintenance ability, material plan, corrosion preventing. It’s helpful to control hazard under accepted level. This method given in the article is a new way to treat HVAC system in pressurized water reactor nuclear power. It hopes that this method will lead to reduce accident loss, save maintenance fee, bring economic benefits and improve the risk of nuclear power.

scholarly journals Transformation of Pressurized Water Reactor (AP1000) to Fuzzy Graph

MATEMATIKA ◽  
2018 ◽  
Vol 34 (2) ◽  
pp. 235-244 ◽  
Author(s):  
Azmirul Ashaari ◽  
Tahir Ahmad ◽  
Wan Munirah Wan Mohamad
Keyword(s):  
Nuclear Power ◽  
Ph Value ◽  
Published Data ◽  
Primary System ◽  
Fuzzy Graph ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fission Process ◽  
Water Reactor ◽  
Dynamic Concentration

Pressurized water reactor (PWR) type AP1000 is a third generation of a nuclear power plant. The primary system of PWR using uranium dioxide to generate heat energy via fission process. The process influences temperature, pressure and pH value of water chemistry of the PWR. The aim of this paper is to transform the primary system of PWR using fuzzy autocatalytic set (FACS). In this work, the background of primary system of PWR and the properties of the model are provided. The simulation result, namely dynamic concentration of PWR is verified against published data.

Study on the Small Pressurized Water Reactor Based on Fully Ceramic Microencapsulated Fuel

2021 ◽  
Author(s):  
Jin Feng Huang
Keyword(s):  
Thermal Conductivity ◽  
Silicon Carbide ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Carbide Matrix ◽  
Silicon Carbide Matrix

Abstract After Fukushima nuclear power plant disaster, the efforts to overcome these defects of PWRs were carried out, such as replacing the cladding and fuel materials. One of these feasible efforts is using Fully Ceramic Microencapsulated (FCM) fuel replacement traditional UO2 pellets fuel into PWR. The FCM fuels are composed of Tri-structural-isotropic (TRISO) particles embedded in silicon carbide matrix. The TRISO fuel can hold its containment integrity and without fission production releases under the design temperature limit of 1600 °C. Furthermore, the silicon carbide matrix will benefit the thermal conductivity, radiation damage resistance, environmental stability and proliferation resistance. Consequently, the safety of the reactor could be significantly improved with FCM fuel instead of the conventional UO2 pellet fuel in PWR. We also analyzed the temperature distribution for the FCM fuel compared the traditional UO2 pellets, the calculation indicated that the centerline temperature is lower than UO2 pellets due to FCM higher thermal conductivity. The calculation demonstrated that, utilizing FCM replacement of conventional UO2 fuel rod is feasible and more security in a small pressurized water reactor. In this paper, a small pressurized water reactor utilizing FCM fuel is considered. A 17 × 17 fuel assemblies with Zircalloy cladding was applied in conceptual design through a preliminary neutronics and thermal hydraulics analysis. The reactor physics is accomplished, including the refuel cycle length, the effective multiplication factor, power distribution analysis being discussed. The Soluble Boron Free (SBF) concepts are introduced in small PWR, as a result, it makes the nuclear power plant more simpler and economical. FCM fuel loading has a very high excess reactivity at the beginning of reactor core life, and it is important to flat reactivity curve during operation, or to minimize excess reactivity during the core life. Consequently, conventional burnable poison configurations were introduced to suppress excess reactivity control at beginning of cycle.

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