scholarly journals Analysis of Steam Generators Corrosion Products from Slovak NPP Bohunice

2012 ◽  
Vol 2012 ◽  
pp. 1-6 ◽  
Author(s):  
Jarmila Degmová ◽  
Július Dekan ◽  
Vladimír Slugeň ◽  
Constanze Thees ◽  
Ivan Smieško ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Construction Materials ◽  
Nuclear Industry ◽  
Feed Water ◽  
Secondary Circuit ◽  
Steam Generators ◽  
Transmission Geometry ◽  
Safety And Reliability ◽  
Main Components

One of the main goals of the nuclear industry is to increase the nuclear safety and reliability of nuclear power plants (NPPs). As the steam generator (SG) is the most corrosion sensitive component of NPPs, it is important to analyze the corrosion process and optimize its construction materials to avoid damages like corrosion cracking. For this purpose two different kinds of SGs and its feed water distributing systems from the NPP Jaslovske Bohunice were studied by nondestructive Mössbauer spectroscopy. The samples were scraped from the surface and analyzed in transmission geometry. Magnetite and hematite were found to be the main components in the corrosion layers of both SGs. Dependant of the material the SG consisted of, and the location in the system where the samples were taken, the ratios between magnetite and hematite and the paramagnetic components were different. The obtained results can be used to improve corrosion safety of the VVER-440 secondary circuit as well as to optimize its water chemistry regime.

scholarly journals Fouling Of Nuclear Steam Generators: Fundamental Studies, Operating Experience and Remedial Measures Using Chemical Additives

2013 ◽  
Vol 2 (1) ◽  
pp. 61-88 ◽  
Author(s):  
C.W. Turner
Keyword(s):  
Economic Performance ◽  
Nuclear Power ◽  
Power Plants ◽  
Tube Bundle ◽  
Operating Experience ◽  
Feed Water ◽  
Chemical Additives ◽  
Steam Generators ◽  
Primary Coolant ◽  
And Performance

Fouling remains a potentially serious issue that if left unchecked can lead to degradation of the safety and performance of nuclear steam generators (SGs). It has been demonstrated that the majority of the corrosion product transported with the feed water to the SGs accumulates in the SG on the tube-bundle. By increasing the risk of tube failure and acting as a barrier to heat transfer, deposit on the tube bundle has the potential to impair the ability of the SG to perform its two safety-critical roles: provision of a barrier to the release of radioactivity from the reactor coolant and removal of heat from the primary coolant during power operation and under certain post accident scenarios. Thus, it is imperative to develop improved ways to mitigate SG fouling for the long-term safe, reliable and economic performance of nuclear power plants (NPPs). This paper provides an overview of our current understanding of the mechanisms by which deposit accumulates on the secondary side of the SG, how this accumulation affects SG performance and how accumulation of deposit can be mitigated using chemical additives to the secondary heat-transport system. The paper concludes with some key questions that remain to be addressed to further advance our knowledge of deposit accumulation and how it can be controlled to maintain safe, economic performance of nuclear SGs.

Comprehensive Analyses of Nuclear Safety System Codes

2013 ◽  
Author(s):  
Glenn A. Roth ◽  
Fatih Aydogan
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Numerical Approach ◽  
Nuclear Industry ◽  
Nuclear System ◽  
Steam Generators ◽  
Two Phase ◽  
Closure Models

Many nuclear system codes have been developed for the main purpose of analyzing reactor performance of a nuclear power plant system during steady state and transient conditions. These codes generally include power plant component models for pumps, pipes, steam generators, pressurizers and other components. The parallel development of these nuclear system codes has been supported by government laboratories, universities, private entities and other organizations throughout the world. This has resulted not only in multiple codes, but multiple versions of the same code with different capabilities. The development paths of each code version have been driven by specific needs. The challenge for the user is to select a code that performs well for the desired analysis problem. Therefore, this work compares different aspects of various nuclear system codes. Firstly, it compares the governing equations for mass, momentum and energy in the evaluated system codes. Secondly, it compares all the codes’ closure models. Closure models are used in system codes to model thermal and mechanical non-equilibrium as well as the coupling of the phases. Thirdly, it compares the Separate Effect Tests (SET) and Integral Effect Tests (IET) employed for the verification and validation (V&V) during the development of each system code. These comparisons cover several thermal and hydraulic models, such as heat transfer coefficients for various flow regimes, two phase pressure correlations, two phase friction correlations, drag coefficients and interfacial models between the fields. Fourthly, major assumptions about the governing and closure equations in these codes are compared and discussed. Fifthly, numerical approach of every code is benchmarked with each other since numerical approach not only affects the speed of the system codes but also the accuracy of the results. Sixthly, the limitations of the codes are evaluated because these codes are challenged by analyzing not only existing nuclear power plants, but also next generation nuclear power plants. The nuclear industry is developing new, innovative reactor designs, such as Small Modular Reactors (SMRs), High-Temperature Gas-cooled Reactors (HTGRs) and others. Sub-types of these reactor designs utilize pebbles, prismatic graphite moderators, helical steam generators, innovative fuel types, and many other design features that may not be fully analyzed by current system codes. The results of this work serve as a guide for development of these system codes and indicate areas where models must be improved to adequately address issues with new reactor design and development activities.

Recent Patents on Valve Mechanism Device

2020 ◽  
Vol 13 (3) ◽  
pp. 230-241
Author(s):  
Ye Dai ◽  
Hui-Bing Zhang ◽  
Yun-Shan Qi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Flow Direction ◽  
Research Progress ◽  
Valve Mechanism ◽  
Advantages And Disadvantages ◽  
Valve Structure ◽  
Valve Leakage ◽  
Safety And Reliability

Background: Valves are an important part of nuclear power plants and are the control equipment used in nuclear power plants. It can change the cross-section of the passage and the flow direction of the medium and has the functions of diversion, cutoff, overflow, and the like. Due to the earthquake, the valve leaks, which will cause a major nuclear accident, endangering people's lives and safety. Objective: The purpose of this study is to synthesize the existing valve devices, summarize and analyze the advantages and disadvantages of various devices from many literatures and patents, and solve some problems of existing valves. Methods: This article summarizes various patents of nuclear-grade valve devices and recent research progress. From the valve structure device, transmission device, a detection device, and finally to the valve test, the advantages and disadvantages of the valve are comprehensively analyzed. Results: By summarizing the characteristics of a large number of valve devices, and analyzing some problems existing in the valves, the outlook for the research and design of nuclear power valves was made, and the planning of the national nuclear power strategic goals and energy security were planned. Conclusion: Valve damage can cause serious safety accidents. The most common is valve leakage. Therefore, the safety and reliability of valves must be taken seriously. By improving the transmission of the valve, the problems of complicated valve structure and high cost are solved.

scholarly journals An Approach to Analyze Diagnosis Errors in Advanced Main Control Room Operations Using the Cause-Based Decision Tree Method

Energies ◽  
2021 ◽  
Vol 14 (13) ◽  
pp. 3832
Author(s):  
Awwal Mohammed Arigi ◽  
Gayoung Park ◽  
Jonghyun Kim
Keyword(s):  
Decision Tree ◽  
Nuclear Power ◽  
Power Plants ◽  
Human Error ◽  
Cognitive Activity ◽  
Emergency Operation ◽  
Nuclear Industry ◽  
Working Environment ◽  
Control Rooms ◽  
Diagnosis Errors

Advancements in the nuclear industry have led to the development of fully digitized main control rooms (MCRs)—often termed advanced MCRs—for newly built nuclear power plants (NPPs). Diagnosis is a major part of the cognitive activity in NPP MCRs. Advanced MCRs are expected to improve the working environment and reduce human error, especially during the diagnosis of unexpected scenarios. However, with the introduction of new types of tasks and errors by digital MCRs, a new method to analyze the diagnosis errors in these new types of MCRs is required. Task analysis for operator diagnosis in an advanced MCR based on emergency operation was performed to determine the error modes. The cause-based decision tree (CBDT) method—originally developed for analog control rooms—was then revised to a modified CBDT (MCBDT) based on the error mode categorizations. This work examines the possible adoption of the MCBDT method for the evaluation of diagnosis errors in advanced MCRs. We have also provided examples of the application of the proposed method to some common human failure events in emergency operations. The results show that with some modifications of the CBDT method, the human reliability in advanced MCRs can be reasonably estimated.

Inservice Testing Program Improvements for New Reactors Small Modular Reactors (SMRs)

2014 ◽  
Author(s):  
Ronald C. Lippy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Small Modular Reactors ◽  
American Society ◽  
Construction Permit ◽  
Regulatory Commission

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed and approved, including the “traditional” reactor designs, the passive safe shutdown designs and the small modular reactors (SMRs). The American Society of Mechanical Engineers (ASME) provides specific Codes used to perform preservice inspection/testing and inservice inspection/testing for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for Design Certifications and Combined Licenses (COLs) under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” in Title 10 of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or following January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME Code for Operation and Maintenance of Nuclear Power Plants (NROMC TG) is assigned the task of ensuring that the preservice testing (PST) and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints (snubbers) in post-2000 nuclear power plants are adequate to provide reasonable assurance that the components will operate as needed when called upon. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in non-safety systems in passive post-2000 reactors including SMRs.

Architecting “Maintenance Program” for Piping and Pipe Support

2013 ◽  
Author(s):  
Koichi Tsumori ◽  
Yoshizumi Fukuhara ◽  
Hiroyuki Terunuma ◽  
Koji Yamamoto ◽  
Satoshi Momiyama
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Industry ◽  
Maintenance Management ◽  
3D Cad ◽  
Cad System ◽  
Nuclear Disaster ◽  
Safety Regulations ◽  
Maintenance Program

A new inspection standard that enhanced quality of operating /maintenance management of the nuclear power plant was introduced in 2009. After the Fukushima Daiichi nuclear disaster (Mar. 11th 2011), the situation surrounding the nuclear industry has dramatically changed, and the requirement for maintenance management of nuclear power plants is pushed for more stringent nuclear safety regulations. The new inspection standard requires enhancing equipment maintenance. It is necessary to enhance maintenance of not only equipment but also piping and pipe support. In this paper, we built the methodology for enhancing maintenance plan by rationalizing and visualizing of piping and pipe support based on the “Maintenance Program” in cooperating with 3D-CAD system.

Research Progress on Radiation Shielding Concrete

2012 ◽  
Vol 253-255 ◽  
pp. 303-307 ◽  
Author(s):  
Jing Yang ◽  
Zhen Fu Chen ◽  
Yuan Chu Gan ◽  
Qiu Wang Tao
Keyword(s):  
Nuclear Waste ◽  
Material Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Radiation Shielding ◽  
Nuclear Industry ◽  
Research Progress ◽  
Concrete Durability ◽  
Technology Research ◽  
Shielding Material

Radiation shielding concrete is widely used in nuclear power plants, accelerators, hospitals, etc. With the development of nuclear industry technology, research on radiation shielding material properties is of great importance. Research on properties of radiation shielding concrete with different aggregates or admixtures and the effect of high temperature on the performance of shielding concrete are introduced. Along with the nuclear waste increase, shielding concrete durability and nuclear waste disposal are getting paramount.

Slow Crack Growth Resistance of Parent and Joint Materials From PE4710 Piping for Safety-Related Nuclear Power Plant Piping

2011 ◽  
Author(s):  
S. Kalyanam ◽  
D.-J. Shim ◽  
P. Krishnaswamy ◽  
Y. Hioe
Keyword(s):  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Slow Crack Growth ◽  
Nuclear Industry ◽  
Pipe Material ◽  
Service Water ◽  
Hdpe Pipe ◽  
Pipe Materials ◽  
Hdpe Pipes

HDPE pipes are considered by the nuclear industry as a potential replacement option to currently employed metallic piping for service-water applications. The pipes operate under high temperatures and pressures. Hence HDPE pipes are being evaluated from perspective of design, operation, and service life requirements before routine installation in nuclear power plants. Various articles of the ASME Code Case N-755 consider the different aspects related to material performance, design, fabrication, and examination of HDPE materials. Amongst them, the material resistance (part of Article 2000) to the slow crack growth (SCG) from flaws/cracks present in HDPE pipe materials is an important concern. Experimental investigations have revealed that there is a marked difference (almost three orders less) in the time to failure when the notch/flaw is in the butt-fusion joint, as opposed to when the notch/flaw is located in the parent HDPE material. As part of ongoing studies, the material resistance to SCG was investigated earlier for unimodal materials. The current study investigated the SCG in parent and butt-fusion joint materials of bimodal HDPE (PE4710) pipe materials acquired from two different manufacturers. The various stages of the specimen deformation and failure during the creep test are characterized. Detailed photographs of the specimen side-surface were used to monitor the specimen damage accumulation and SCG. The SCG was tested using a large specimen (large creep frame) as well as using a smaller size specimen (PENT frame) and the results were compared. Further, the effect of polymer orientation or microstructure in the bimodal HDPE pipe on the SCG was studied using specimens with axial and circumferential notch orientations in the parent pipe material.

scholarly journals Calculation of Potential Radiation Doses Associated with Predisposal Management of Dismantled Steam Generators from Nuclear Power Plants

2020 ◽  
Vol 12 (12) ◽  
pp. 5149
Author(s):  
Ga Hyun Chun ◽  
Jin-ho Park ◽  
Jae Hak Cheong
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Radiation Doses ◽  
Steam Generators ◽  
Pressurized Water ◽  
Water Reactor ◽  
General Pathway ◽  
Heavy Water Reactor ◽  
Past Experiences

Although the generation of large components from nuclear power plants is expected to gradually increase in the future, comprehensive studies on the radiological risks of the predisposal management of large components have been rarely reported in open literature. With a view to generalizing the assessment framework for the radiological risks of the processing and transport of a representative large component—a steam generator—12 scenarios were modeled in this study based on past experiences and practices. In addition, the general pathway dose factors normalized to the unit activity concentration of radionuclides for processing and transportation were derived. Using the general pathway dose factors, as derived using the approach established in this study, a specific assessment was conducted for steam generators from a pressurized water reactor (PWR) or a pressurized heavy water reactor (PHWR) in Korea. In order to demonstrate the applicability of the developed approach, radiation doses reported from actual experiences and studies are compared to the calculated values in this study. The applicability of special arrangement transportation of steam generators assumed in this study is evaluated in accordance with international guidance. The generalized approach to assessing the radiation doses can be used to support optimizing the predisposal management of large components in terms of radiological risk.

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