Pump and Valve Fastener Serviceability in PWR Nuclear Facilities

1996 ◽  
Vol 118 (1) ◽  
pp. 38-41
Author(s):  
N. T. Moisidis ◽  
M. D. Ratiu
Keyword(s):  
Boric Acid ◽  
Power Plants ◽  
Selection Criterion ◽  
Low Alloy Steels ◽  
Pressurized Water Reactor ◽  
Nuclear Facilities ◽  
Pressurized Water ◽  
Alloy Steels ◽  
Borated Water ◽  
Maintenance Problem

The results of several studies conducted on corrosion of carbon and low-alloy steels in borated water have shown that impingement of borated steam on ferritic steels or contact with a moist paste of boric acid can lead to high corrosion rates due to high local concentrations of boric acid on the surface. The corrosion process of the flange fasteners of pumps and valves is considered a material compatibility and equipment maintenance problem. Therefore, the nuclear utilities of pressurized water reactor (PWR) power plants can prevent this damage by implementing appropriate fastener steel replacement and extended inspections to detect and correct the cause of leakage. A 3-phase corrosion protection program is presented for implementation based on system operability, outage-related accessibility, and cost of fastener replacement versus maintenance frequency increase. A selection criterion for fastener material is indicated based on service limitation: preloading and metal temperature.

Boric Acid Corrosion in the Primary System of PWRs

2014 ◽  
Vol 1081 ◽  
pp. 58-62 ◽  
Author(s):  
Tian Li ◽  
Zhi Zhen Peng ◽  
Yan Cheng Fan ◽  
Xi Ming Tang ◽  
Chen Guang Kong ◽  
...  
Keyword(s):  
Boric Acid ◽  
Nuclear Power ◽  
Power Plants ◽  
Visual Inspection ◽  
Acid Corrosion ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Water Leakage ◽  
Pressurized Water ◽  
Borated Water

Borated water leakage could lead to boric acid corrosion of nuclear equipment, and it’s a danger to the pressurized water reactor nuclear power plants (PWRs). In this study, the principle of boric acid corrosion was studied first. Through the experiences of borated water leakages and boric acid corrosion at home and abroad, the main leakage sources in PWRs were summarized. Combining the technique features of visual inspection, the detecting principles were proposed. In the end, a protection strategy of boric acid corrosion was established including the control of borated water leakage, the detection of equipment, the treatment of leakages and the improvement of the equipment material.

The Failure Analysis and Processing of Digital Reactor Protection System

2013 ◽  
Author(s):  
Jun Zhao ◽  
Xing Zhou ◽  
Jin Hu ◽  
Yanling Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Phase 1 ◽  
Protection System ◽  
Pressurized Water Reactor ◽  
Communication Failure ◽  
Pressurized Water ◽  
Power Failure ◽  
Self Test ◽  
Reactor Protection System

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1. The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security. There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system. The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1. It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.

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.

Environmental Fatigue Test of CF8M With a Small Autoclave Simulating PWR Conditions

2008 ◽  
Author(s):  
Il-Seok Jeong ◽  
Gag-Hyeon Ha ◽  
Tae-Ryoung Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Low Cycle Fatigue ◽  
Test Procedure ◽  
Cyclic Strain ◽  
Fatigue Tests ◽  
Operating Conditions ◽  
Measuring Instruments ◽  
Pressurized Water Reactor ◽  
Pressurized Water

To develop a fatigue design curve of cast stainless steel CF8M used in primary piping material of nuclear power plants, low-cycle fatigue tests have been conducted by Korea Electric Power Research Institute (KEPRI). A small autoclave simulated the environment of a pressurized water reactor (PWR), 15 MPa and 315 °C. Fatigue life was measured in terms of the number of cycles with the variation of strain amplitudes at 0.04%/s strain rate. A small autoclave of 1 liter and cylindrical solid fatigue specimens were used for the strain-controlled low cycle environmental fatigue tests to make the experiments convenient. However, it was difficult to install displacement measuring instruments at the target length of the specimens inside the autoclave. To mitigate the difficulty displacement data measured at the shoulders of the specimen were calibrated based on the data relation of the target and shoulder length of the specimen during hot air test conditions. KEPRI developed a test procedure to perform low cycle environmental fatigue tests in the small autoclave. The procedure corrects the cyclic strain hardening effect by performing additional tests in high temperature air condition. KEPRI verified that the corrected test result agreed well with that of finite element method analysis. The process of correcting environmental fatigue data would be useful for producing reliable fatigue curves using a small autoclave simulating the operating conditions of a PWR.

Mechanical Properties of SA508 Gr.4N Model Alloys as a High Strength RPV Steel

2010 ◽  
Author(s):  
Min-Chul Kim ◽  
Ki-Hyoung Lee ◽  
Bong-Sang Lee ◽  
Whung-Whoe Kim
Keyword(s):  
Mechanical Properties ◽  
Fracture Toughness ◽  
Impact Toughness ◽  
Power Plants ◽  
Charpy Impact ◽  
Alloying Elements ◽  
Low Alloy Steels ◽  
Charpy Impact Toughness ◽  
Age Related ◽  
Alloy Steels

Demands of RPV materials with higher strength and toughness are rising to increase the power capacity and the operation life of nuclear power plants. The ASME SA508 Gr.4N specification can give a superior toughness and strength to the commercial low alloy steels such as SA508 Gr.3. However, the SA508-Gr.4N steels have not yet been used commercially due to a lack of information of the productivity and the age related properties. While the irradiation embrittlement studies are going-on, the current paper focused on the effects of alloying elements such as Ni, Cr and Mo on the fracture mechanical properties of the SA508 Gr.4N low alloy steels. Various model alloys were fabricated by changing the contents of alloying elements based on the composition range of the ASME specification. Tensile properties, Charpy impact toughness and fracture toughness of the model alloys were evaluated and those properties were discussed with the microstructural characteristics of each alloy. The strengths of the alloys were increased with increase of the Ni and Mo contents while there was no remarkable change of the yield strength with the Cr addition. The Charpy impact and fracture toughness were considerably improved with the increase of Ni, Cr contents. The Mo addition did not change the toughness properties significantly. The Cr contents were more effective on the fracture toughness through changing the carbides precipitation characteristics and the Ni contents were effective on the Charpy impact toughness through changing the effective grain size.

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.

scholarly journals Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2

2021 ◽  
Vol 13 (14) ◽  
pp. 7964
Author(s):  
Alain Flores y Flores ◽  
Danilo Ferretto ◽  
Tereza Marková ◽  
Guido Mazzini
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Nuclear Industry ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Accident Simulation ◽  
Experimental Facilities ◽  
The Difference

The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects.

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