Over Limit Oxygen Effect on Crack Propagation of WWER 440 RPV Steel in Primary Circuit Water

2009 ◽  
Author(s):  
Miroslava Ernestova ◽  
Anna Hojna
Keyword(s):  
Crack Propagation ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Environment ◽  
Nuclear Research ◽  
Compact Tension ◽  
Oxygen Effect ◽  
Primary Circuit ◽  
Nuclear Research Institute ◽  
Primary Water

Experience with operating nuclear power plants worldwide reveals that many failures may be attributed to fatigue associated with mechanical loading due to vibration and with corrosion effect due to exposure to high-temperature environment. In order to clarify the simultaneous influence on reactor pressure vessel (RPV) material testing of ferritic steel 15Ch2MFA used for RPV of WWER 440 was performed at Nuclear Research Institute (NRI) autoclaves. Cyclic and constant loadings were applied to Compact Tension (CT) specimens in WWER primary water environment at 290°C and simultaneous effect of different oxygen levels (< 20 ppb, 200 ppb, 2000 ppb) on crack propagation has been evaluated. Obtained crack growth rates are compared with ASME XI Code and VERLIFE curves and crack behaviour is discussed.

Decommissioning of Obsolete Nuclear Facilities in the Nuclear Research Institute Rez

2011 ◽  
Author(s):  
Josef Podlaha ◽  
Karel Svoboda ◽  
Eduard Hansli´k
Keyword(s):  
Czech Republic ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Research ◽  
Current Status ◽  
Joint Stock Company ◽  
Nuclear Facilities ◽  
Nuclear Research Institute ◽  
The Czech Republic ◽  
Research Institute

After more than 55 years of activities of the Nuclear Research Institute Rez (NRI) in the nuclear field, there are some obsolete nuclear facilities that shall be decommissioned. NRI is a leading institution in all areas of nuclear R&D in the Czech Republic. NRI has had a dominant position in the nuclear programme since it was established in 1955 as a state-owned research organization and it has developed to its current status. In December 1992, NRI has been transformed into a joint-stock company. The Institute’s activity encompasses nuclear physics, chemistry, nuclear power, experiments at the research reactor and many other topics. Main issues addressed in NRI in the past decades were concentrated on research, development and services provided to the nuclear power plants operating VVER reactors, development of chemical technologies for fuel cycle and irradiation services to research and development in the industrial sector, agriculture, food processing and medicine. The NRI operates two research nuclear reactors, many facilities as a hot cell facility, research laboratories, technology for radioactive waste (RAW) management, radionuclide irradiators, an electron accelerator, etc. The obsolete facilities to be decommissioned comprise various research facilities and facilities for RAW management. Decommissioning of nuclear facilities NRI is the only ongoing decommissioning project in the Czech Republic. Decommissioning started in 2003 and will be finished in 2014. Some facilities have already been successfully decommissioned.

Verification Analyses for Newly Developed Automatic 3D Finite Element Crack Propagation System

2014 ◽  
Author(s):  
Hiroaki Doi ◽  
Hitoshi Nakamura ◽  
Wenwei Gu ◽  
Do-Jun Shim ◽  
Gery Wilkowski
Keyword(s):  
Finite Element ◽  
Crack Propagation ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
3D Finite Element ◽  
Primary Water ◽  
Axial Crack ◽  
Analysis System

In order to calculate the crack propagation in complicated-shaped locations in components such as weld in penetration structures of reactor pressure vessel of nuclear power plants, an automatic 3D finite element crack propagation system (CRACK-FEM) has been developed by the Nuclear Regulation Authority (NRA, Japan). To confirm the accuracy and effectiveness of this analysis system, a verification analysis was performed. The program used for comparison is PipeFracCAE developed by Engineering Mechanics Corporation of Columbus, which has been used for many crack propagation analyses in various applications. In this paper, the axial crack propagation analysis for primary water stress corrosion cracking (PWSCC) in a steam generator inlet nozzle of a pressurized water reactor (PWR) plant is presented. The results demonstrate that the two codes are in good agreement. The contents of this paper were conducted as a research project of the Japan Nuclear Energy Safety Organization (JNES) when one of the authors (Doi) belongs to JNES. After this project, JNES was abolished and its staff and task were absorbed into NRA on March 1, 2014.

Development of a New Specimen to Study Crack Propagation Threshold and Non-Propagation Conditions

2011 ◽  
Author(s):  
Pauline Bouin ◽  
Ste´phane Marie ◽  
Gre´gory Perez
Keyword(s):  
Crack Propagation ◽  
Nuclear Power ◽  
Power Plants ◽  
Complex Loading ◽  
Crack Arrest ◽  
Compact Tension ◽  
Element Analysis ◽  
Loading Ratio ◽  
Nuclear Structures ◽  
Catastrophic Fracture

The extension of nominal service-life of nuclear power plants leads to calculate more involved safety analyses of the nuclear structures for the different scenario of failures. In order to prevent catastrophic fracture, it is important to minimise the initiation of cracks or if considering that a crack can initiate, to determine if the extension of these cracks can stop or lead to the failure of the structure integrity. As a result, this work requires better understanding and characterisation of the crack propagation or non-propagation behaviour. Existing practises to determinate fatigue crack growth rate and fatigue thresholds are mainly based on compact tension (CT) specimen testing. For high loading ratio, the maximum of stress intensity factor (SIF) usually remains constant whereas the minimum of the SIF is increased to reduce the SIF range down to the threshold. On the contrary, for low loading ratio, complex loading control is required to make maximum and minimum SIF decrease. Furthermore, the results of these tests are mainly dependent on the accuracy of the piloting of the test bench, the specimen instrumentation and the force loading cell capacity. This paper presents details of adaptation of a specimen initially developed to study crack arrest problems under cleavage fracture. A bulk compact tension (BCT) specimen has been designed based on a standard CT specimen. Specifically, a reinforced heel at the back of the specimen enables a reduction of the loading at the crack front. In the case of fatigue, this reduces the SIF range while the loading condition remains constant during the test. A SIF calibration for these BCT specimens has also been established to estimate the compliance using finite element analysis with the French code, Cast3M.

An Efficient, Flexible Arrangement to Generate High Quality Process or Domestic Heat in Recuperative Gas Cycle Power Plants: Application to Helium Direct Cycle Nuclear Power Plants

1979 ◽  
Vol 101 (1) ◽  
pp. 130-140 ◽  
Author(s):  
Z. P. Tilliette ◽  
B. Pierre ◽  
P. F. Jude
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Hot Water ◽  
Primary Circuit ◽  
Inlet Temperature ◽  
Total Efficiency ◽  
High Quality ◽  
Quality Process ◽  
Direct Cycle

The advantages of gas turbine power plants in general and closed cycle systems under gas pressure in particular for waste heat recovery are well known. A satisfactory efficiency for electric power generation and good conditions to obtain a significant amount of hot water above 100°C lead to a high fuel utilization. However, as in most of projects, it is not much possible to produce high temperature steam or water without significantly decreasing the electricity production. A new method for an additional generation of high quality process or domestic heat is proposed. The basic feature of this method lies in arranging one or two steam generators or preheaters in parallel with the low pressure side of the recuperator. The high total efficiency and the noteworthy flexibility of this system are emphasized. This arrangement is suitable for any kind of heat source, but the applications presented in this paper are related to helium direct cycle nuclear power plants the main features of which are a single 600 MW(e) turbomachine, a turbine inlet temperature of 775°C, no or one intermediate cooling and a primary circuit fully integrated in a pre-stressed concrete reactor vessel.

Economic Evaluation of Proactive Replacement of Upper Internals Guide Tube Support Pins (Split Pins)

2005 ◽  
Author(s):  
Robert K. Perdue ◽  
G. Gary Elder ◽  
Gregory Gerzen
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Net Present Value ◽  
Time To Failure ◽  
Guide Tube ◽  
Reliability Modeling ◽  
Nuclear Plants ◽  
Zinc Addition ◽  
Primary Water ◽  
The Impact

Certain nuclear power plants have “Rev B” reactor vessel upper internals guide tube support pins, commonly referred to as split pins, made from material with properties similar to Alloy 600 and known to be susceptible to primary water stress corrosion cracking (PWSCC). This paper describes a rigorous probabilistic methodology for evaluating the economics of a preemptive replacement of these split pins, and describes an application at four of Exelon Generation’s nuclear plants. The method uses Bayesian statistical reliability modeling to estimate a Weibull time-to-failure prediction model using limited historical failures, and a Westinghouse proactive aging management simulation tool called PAM to select a split pin replacement date that would maximize the net present value of cash flow to a plant. Also in this study is a sensitivity evaluation of the impact of zinc addition on split pin replacement timing. Plant decisions made based in part on results derived from applying this approach are noted.

Overview of Recent NUGENIA Activities Particularly in Relation to Structural Integrity

2015 ◽  
Author(s):  
John Sharples ◽  
Elisabeth Keim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Nuclear Research ◽  
Joint Research ◽  
Non Profit ◽  
Integrity Assessment ◽  
Industry Research ◽  
Implementation And Dissemination ◽  
Technical Area

NUGENIA, an international non-profit association founded under Belgian legislation and launched in March 2012, is dedicated to nuclear research and development (R&D) with a focus on Generation II and III power plants. NUGENIA is the integrated framework between industry, research and safety organisations for safe, reliable and competitive nuclear power production, and is aimed at running an open innovation marketplace, to promote the emergence of joint research and to facilitate the implementation and dissemination of R&D results. The technical scope of NUGENIA consists of eight technical areas. One of these areas, Technical Area 4, is associated with the structural integrity assessment of systems, structures and components. A brief overview of recent NUGENIA activities in general is provided in this paper and a specific focus is given on developments in relation to Technical Area 4.

Environmental Assisted Fatigue and EDF 900 MWe PWRs Fleet: Towards an Exemption of Environmental Effects Consideration for Secondary Circuit Components

2018 ◽  
Author(s):  
Sam Cuvilliez ◽  
Gaëlle Léopold ◽  
Thomas Métais
Keyword(s):  
Environmental Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Fatigue Loading ◽  
Degradation Process ◽  
Fatigue Curve ◽  
Primary Circuit ◽  
Secondary Circuit ◽  
Technical Specifications ◽  
Circuit Components

Environmentally Assisted Fatigue (EAF) is receiving nowadays an increased level of attention for existing Nuclear Power Plants (NPPs) as utilities are now working to extend their life. In the wake of numerous experimental fatigue tests carried out in air and also in a PWR environment, the French RCC-M code [1] has recently been amended (in its 2016 edition) with two Rules in Probatory Phase (RPP), equivalent to ASME code-cases, “RPP-2” and “RPP-3” [2] [3]. RPP-2 consists of an update of the design fatigue curve in air for stainless steels (SSs) and nickel-based alloys, and is also associated with RPP-3 which provides guidelines for incorporating the environmental penalty “Fen” factor in fatigue usage factor calculations. Alongside this codification effort, an EAF screening has recently been carried out within EDF DT [4] on various areas of the primary circuit of the 900 MWe plants of the EDF fleet. This screening led to the identification of a list of 35 “sentinel locations” which are defined as areas most prone to EAF degradation process. These locations will be subjected to detailed EAF analysis in the stress report calculations (according to the above-mentioned RCC-M code cases) for the fourth decennial inspection of the 900 MWe (VD4 900 MWe) power plants. The potential impact of EAF on the secondary circuit components is another question to address in anticipation of the VD4 900 MWe, as they may be considered as class 1 or class 2 equipment for RCC-M application according to the equipment specification. This paper presents the approach proposed by EDF towards an exemption of environmental effects consideration for secondary circuit components. The argument is first based on a review of experimental campaigns led in Japan and France (respectively on fatigue test specimens and at the component scale) which indicate a Dissolved Oxygen (DO) content threshold below which environmental effects are almost inexistent. The (conservative) value of 40 ppb has been selected consistently with NUREG/CR-6909 revision 0 [5]. The second part of the argument is built, on the one hand, on the analysis of the EDF Technical Specifications for Operation (STE) which narrows the scope of the study only to unit outages, and, on the other hand, on the analysis of 5 years of operations of all 900 MWe plants of the EDF fleet (equivalent to 170 reactor-years). It has been shown that the DO content rarely exceeded the 40 ppb threshold in the secondary coolant, and that in this case, the considered locations were not submitted to any fatigue loading.

Research on the Micro-Mechanical State at Tip of Environmentally Assisted Cracking Based on Latin Hypercube Sampling Method

2019 ◽  
Vol 795 ◽  
pp. 74-78
Author(s):  
Kuan Zhao ◽  
He Xue ◽  
Ling Yan Zhao
Keyword(s):  
Oxide Film ◽  
Material Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Sampling Method ◽  
Latin Hypercube Sampling ◽  
Primary Circuit ◽  
Mechanical State ◽  
Latin Hypercube ◽  
Environmentally Assisted Cracking

Environmentally assisted cracking (EAC) of nickel-based alloys is one of the most significant potential safety hazards in the primary circuit of nuclear power plants. To understand the influence of randomness on micro-mechanical state at tip of EAC, Latin hypercube sampling method is applied to analyze the uncertainty of stress-strain in the oxide film at the EAC tip considering the uncertainties of load and material properties of base metal and oxide film. Meanwhile, to improve the efficiency of numerical analysis, MATLAB is employed in the secondary development for ABAQUS. With the help of finite element numerical simulation and Latin hypercube sampling method, the uncertainty of mechanical properties at tip of EAC in one-inch compact tension specimen is simulated and analyzed in this study. The results show that the randomness of material properties and load markedly affect the uncertainty of micro-mechanical state. Among the variables, The randomness of load has the greatest influence on uncertainty of strain, and Poisson`s ratio of oxide film is the smallest effect.

PWSCC Crack Growth Mitigation With Inlay

2011 ◽  
Author(s):  
David Rudland ◽  
Frederick W. Brust ◽  
D. J. Shim ◽  
G. Wilkowski
Keyword(s):  
Residual Stresses ◽  
Weld Metal ◽  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Resistant Material ◽  
Residual Stress State ◽  
Weld Material ◽  
Primary Water ◽  
Mitigation Methods

Primary water stress corrosion cracking (PWSCC) is an issue of concern in the dissimilar metal welds (DMW) connecting vessel nozzles and stainless steel piping in PWR nuclear power plants. PWSCC occurs due to the synergistic interaction of several factors including tensile weld residual stresses, a corrosion sensitive weld metal (usually Alloy 82/182 weld metal) and a corrosive environment. Several mechanical mitigation methods to control PWSCC have been developed in order to alter the weld residual stresses on the nozzle. These methods consist of applying a weld overlay repair (WOR), using a method called mechanical stress improvement process (MSIP), and applying an inlay to the nozzle ID, the latter of which is the subject of this paper. An inlay consists of machining the pipe ID at the region of the DMW and applying a PWSCC resistant weld material at the machined region. The PWSCC resistant material is mainly Alloy 52/152, which has a higher chromium content compared with Alloy 82/182. The inlay is a corrosion resistant material, and the proposed application thickness (after final machining) is 3 mm. Therefore, once the crack grows through the inlay, the growth in the underlying A82/182 material is much faster. This leads to a complicated crack shape which is small at the nozzle ID and becomes larger in the original weld material and approaches a balloon shape. Here the weld residual stress state caused by the inlay is first discussed. Next, the effect of crack growth through the inlay and into the underlying Alloy 82/182 material is discussed. Finally, implications of inlay for mitigation and consideration of alternatives is discussed.

Laser-Based Maintenance and Repair Technologies for Reactor Components

2004 ◽  
Author(s):  
Masaki Yoda ◽  
Naruhiko Mukai ◽  
Makoto Ochiai ◽  
Masataka Tamura ◽  
Satoshi Okada ◽  
...  
Keyword(s):  
Optical Fiber ◽  
Preventive Maintenance ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Rapid Heating ◽  
Water Environment ◽  
Inspection System ◽  
Material Surface ◽  
Maintenance And Repair

Stress corrosion cracking (SCC) is the major factor to reduce the reliability of aged reactor components. Toshiba has developed various laser-based maintenance and repair technologies and applied them to existing nuclear power plants. Laser-based technology is considered to be the best tool for remote processing in nuclear power plants, and particularly so for the maintenance and repair of reactor core components. Accessibility could be drastically improved by a simple handling system owing to the absence of reactive force against laser irradiation and the flexible optical fiber. For the preventive maintenance, laser peening (LP) technology was developed and applied to reactor components in operating BWR plants. LP is a novel process to improve residual stress from tensile to compressive on material surface layer by irradiating focused high-power laser pulses in water. We have developed a fiber-delivered LP system as a preventive maintenance measure against SCC. Laser ultrasonic testing (LUT) has a great potential to be applied to the remote inspection of reactor components. Laser-induced surface acoustic wave (SAW) inspection system was developed using a compact probe with a multi-mode optical fiber and an interferometer. The developed system successfully detected a micro slit of 0.5mm depth on weld metal and heat-affected zone (HAZ). An artificial SCC was also detected by the system. We are developing a new LP system combined with LUT to treat the inner surface of bottom-mounted instruments (BMI) of PWR plants. Underwater laser seal welding (LSW) technology was also developed to apply surface crack. LSW is expected to isolate the crack tip from corrosive water environment and to stop the propagation of the crack. Rapid heating and cooling of the process minimize the heat effect, which extends the applicability to neutron-irradiated material. This paper describes recent advances in the development and application of such laser-based technologies.

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