Homogenization of Cr-Ni Austenitic Steel Studying: Liquation and Microhardness Heterogeneity Equalization

2019 ◽  
Vol 822 ◽  
pp. 53-59 ◽  
Author(s):  
Anton Sergeevich Tsvetkov ◽  
Irina Vladimirovna Teplukhina
Keyword(s):  
Stainless Steel ◽  
Service Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Diffusion Experiment ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Exposure Effect ◽  
Alloy Type ◽  
Radiation Swelling

Currently, a new generation of pressurized water reactor for nuclear power plants with an extended service life (up to 60 years) and a guarantee of their complete safety are being designed in Russia. Analysis of the reactor internal elements performance showed, that designed service life cannot be guaranteed if the reactor’s internal parts would be made from currently used stainless steel (18-10 alloy type). Instead of the used steel, to ensure operability, new austenitic stainless steel (16-25 alloy type), with increased resistance to radiation swelling, is being developed for production of forged ring blanks for core baffle. The use of new steel requires revision of the existing metallurgical production technology stages. Therefore in this paper diffusion experiment was carried out to determine the required duration of homogenization. The results are presented in terms of different duration of the high-temperature exposure effect on the liquation heterogeneity equalization. The relation between duration of homogenization and microhardness is also shown.

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.

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.

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.

Parametric Study on Determining Optimal Eddy Current Testing Method for Inspection of Steam Generator Tubes in SMART Plant

2010 ◽  
Author(s):  
Dae-Kwang Kim ◽  
Sung-Jin Han ◽  
Hak-Joon Kim ◽  
Sung-Jin Song ◽  
Yun-hang Choung
Keyword(s):  
Steam Generator ◽  
Eddy Current ◽  
Nuclear Power ◽  
Power Plants ◽  
Eddy Current Testing ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Current Testing ◽  
Current Test ◽  
Testing Condition

The SMART (System-integrated Modular Advanced ReacTor) is small sized integral type pressurized water reactor designed by KAERI (Korea Atomic Energy Research Institute), Korea. But, shape of steam generator (SG) in SMART plant differs from those in operated nuclear power plants (NPPs). Especially, SG tubes in SAMRT plant is helical type with around 600 mm of innermost diameter and thickness of 2.5 mm which is thicker than general NPPs one. For providing integrity of SG tube in SMART plant, new types of ECT method are needed because eddy current testing (ECT) is one of widely adopted method for inspection of SG tubes in NPPs. Therefore, in this study, we investigate optimal conditions or parameters for detecting and evaluating of flaws in the SG tubes in SMART plant by simulation of ECT signals with various testing condition or parameter such as frequency, coil gap and etc. From the simulated ECT signals optimal eddy current test condition or parameters are proposed.

Simulation of Triaxial Residual Stress Mapping for a Hollow Cylinder

2012 ◽  
Author(s):  
Mitchell D. Olson ◽  
Wilson Wong ◽  
Michael R. Hill
Keyword(s):  
Residual Stress ◽  
Stainless Steel ◽  
Nuclear Power ◽  
Complex Geometry ◽  
Cylindrical Body ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Weld Residual Stress ◽  
Novel Method ◽  
Stress Mapping

This paper describes a novel method to determine a two-dimensional map of the triaxial residual stress on a radial-axial plane of interest in a hollow cylindrical body. With the description in hand, we present a simulation to validate the steps of the method. The simulation subject is a welded cylindrical nozzle typical of a nuclear power pressurized water reactor pressurizer; in the weld region, the nozzle inner diameter is roughly 132 mm (5.2 inch) and the wall thickness is roughly 35 mm (1.4 inch). The pressure vessel side of the nozzle is carbon steel (with a thin stainless steel lining), the piping side is austenitic stainless steel, and between the two are weld and buttering deposits of nickel alloy. Weld residual stresses in such nozzles have important effects on crack growth rates in fatigue and stress corrosion cracking, therefore measurements of weld residual stress can help provide inputs for managing aging reactor fleets. Nuclear power plant welds often have large and complex geometry, which has made residual stress measurements difficult, and this work provides a proof of concept for a new experimental technique for measurements on welded nozzles.

Age-Related Degradation of Steam Generator Internals Based on Industry Responses to Generic Letter 97-06

2002 ◽  
Author(s):  
M. Subudhi ◽  
E. J. Sullivan
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressurized Water Reactor ◽  
Long Term Effects ◽  
Steam Generators ◽  
Pressurized Water ◽  
Age Related ◽  
The U.S

This paper presents the results of an aging assessment of the nuclear power industry’s responses to NRC Generic Letter 97-06 on the degradation of steam generator internals experienced at Electricite de France (EdF) plants in France and at a United States pressurized water reactor (PWR). Westinghouse (W), Combustion Engineering (CE), and Babcock & Wilcox (B & W) steam generator models, currently in service at U.S. nuclear power plants, potentially could experience degradation similar to that found at EdF plants and the U.S. plant. The steam generators in many of the U.S. PWRs have been replaced with steam generators with improved designs and materials. These replacement steam generators have been manufactured in the U.S. and abroad. During this assessment, each of the three owners groups (W, CE, and B&W) identified for its steam generator models all the potential internal components that are vulnerable to degradation while in service. Each owners group developed inspection and monitoring guidance and recommendations for its particular steam generator models. The Nuclear Energy Institute incorporated in NEI 97-06, “Steam Generator Program Guidelines,” a requirement to monitor secondary side steam generator components if their failure could prevent the steam generator from fulfilling its intended safety-related function. Licensees indicated that they implemented or planned to implement, as appropriate for their steam generators, their owners group recommendations to address the long-term effects of the potential degradation mechanisms associated with the steam generator internals.

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