Flaw Acceptance Rule of J-Groove Weld of Bottom Mounted Instrumentation Nozzle

2014 ◽  
Author(s):  
Kiminobu Hojo ◽  
Yasuto Nagoshi ◽  
Mayumi Ochi ◽  
Naoki Miura ◽  
Masayuki Kamaya ◽  
...  
Keyword(s):  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Fe Analysis ◽  
Evaluation Procedure ◽  
Weld Region ◽  
Acceptance Rule ◽  
Nuclear Plants ◽  
Simple Geometry ◽  
Calculation Results

The rules on fitness for service for nuclear power plants of JSME are applied for flaw evaluation after detecting defects in the operating nuclear plants in Japan. The rules mainly focus on simple geometry such as straight pipes or vessels and do not provide the evaluation procedure for complex structures. The authors made a draft of flaw acceptance rule for J-groove weld of a bottom mounted instrumentation nozzle at application of the cap repair. The rule contains flaw modeling, fatigue and SCC crack growth calculation and flaw instability assessment. After detecting a defect on a J-groove weld, a flaw will be modeled in the whole of J-groove weld region because of fast SCC crack propagation in the weld region. Due to complex configuration of the evaluation location, FE analysis is needed for obtaining stress intensity factors (Ks) to calculate the crack growth and flaw instability. The proposed rule has a guidance for K calculation by FE analysis with the aim of decreasing dependence of individuals for calculation. The authors performed benchmark analyses to confirm the guidance applicability. The calculation results by three participants agreed within several percent.

Application of Stress Intensity Factors for Deep Surface Cracks to Crack Growth Evaluation

2013 ◽  
Author(s):  
Fuminori Iwamatsu ◽  
Katsumasa Miyazaki ◽  
Hajime Miyata ◽  
Hideki Yuya
Keyword(s):  
Aspect Ratio ◽  
Crack Growth ◽  
Surface Crack ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Procedure ◽  
Crack Growth Behavior ◽  
Surface Cracks ◽  
Deep Surface ◽  
Growth Evaluation

Flaw evaluation for nuclear power plants is conducted on the basis of a fitness-for-service code. For instance, the ASME Boiler and Pressure Vessel Code Section XI (ASME Section XI) and the JSME Rules on Fitness-for-Service for Nuclear Power Plants (JSME Code) prescribe a flaw evaluation procedure. In flaw evaluation, an aspect ratio of a detected surface crack is defined by a/l, where a is the crack depth and l is the crack length, and the aspect ratio a/l does not exceed 0.5. Therefore, a deep surface crack, which has an aspect ratio a/l greater than 0.5, is characterized as a semicircle with l = 2a. Meanwhile, deep surface cracks caused by stress corrosion cracking (SCC) have been detected in the Ni based alloy weld metal. Since the limit of the flaw characterization rule which is an aspect ratio a/l ≤ 0.5 seems to conduct to a conservative evaluation result for a deep surface crack, more rational and applicable flaw evaluation is required in order to eliminate surplus conservatism. In this study, a flaw evaluation procedure based on ASME Section XI or JSME Code is extended to deal with a deep surface crack. To evaluate crack growth behavior for a deep surface crack, coefficients to calculate stress intensity factors were evaluated by finite element analysis (FEA) and shown in tabular form on the basis of equations prescribed in ASME Section XI and JSME Code. To verify the applicability of proposed coefficients to crack growth evaluation, SCC crack growth behavior for a deep initial crack was evaluated by coefficients applied to the ASME Section XI procedure and a detailed FEA method. Applicability of coefficients to crack growth evaluation was verified through comparisons of crack growth behaviors for deep surface crack under various stress fields.

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 Application of Explosion-Proof RFID Technology in Nuclear Power Plants

2020 ◽  
Vol 20 (2) ◽  
pp. 127-132
Author(s):  
Namjin Cho ◽  
Dongsu Im ◽  
Jungdon Kwon ◽  
Teayeon Cho ◽  
Junglim Lee
Keyword(s):  
Radio Frequency ◽  
Radio Frequency Identification ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Ignition Source ◽  
Rfid Technology ◽  
Nuclear Plants ◽  
Frequency Identification

Nuclear power plants store and use flammable gases and liquids and consequently risk explosions. Therefore, nuclear plants employ explosion-proof equipment; however, this equipment is not always sufficiently maintained. This lack of maintenance can affect the safety-related equipment intended to shut down the reactor, because the explosion-proof equipment itself can act as an ignition source. Radio-frequency identification (RFID) technology should be explored as a tool to improve both the convenience and efficiency of maintenance. We analyzed and compared explosion-proof RFID technology that can be used in nuclear power plants.

scholarly journals Effects of different nuclear evaluation data on the RMC keff calculation

2020 ◽  
Vol 239 ◽  
pp. 19005
Author(s):  
Zhang Wenxin ◽  
Qiang shenglong ◽  
Yin qiang ◽  
Cui Xiantao
Keyword(s):  
Cross Section ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Section Data ◽  
Calculation Results ◽  
The Impact ◽  
Physical Calculation

Neutron cross section data is the basis of nuclear reactor physical calculation and has a decisive influence on the accuracy of calculation results. AFA3Gassemble is widely used in nuclear power plants. CENACE is an ACE format multiple-temperature continuous energy cross section library that developed by China Nuclear Data Centre. In this paper, we calculated the AFA3G assemble by RMC.We respectively used ENDF6.8/, ENDF/7 and CENACE data for calculation. The impact of nuclear data on RMC calculation is studied by comparing the results of different nuclear data.

scholarly journals Interaction of Cyclic Loading (Low-Cyclic Fatigue) with Stress Corrosion Cracking (SCC) Growth Rate

2020 ◽  
Vol 2020 ◽  
pp. 1-10
Author(s):  
Rehmat Bashir ◽  
He Xue ◽  
Rui Guo ◽  
Yueqi Bi ◽  
Muhammad Usman
Keyword(s):  
Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Cyclic Fatigue ◽  
Joint Effect ◽  
Corrosion Cracking

The structural integrity analysis of nuclear power plants (NPPs) is an essential procedure since the age of NPPs is increasing constantly while the number of new NPPs is still limited. Low-cyclic fatigue (LCF) and stress corrosion cracking (SSC) are the two main causes of failure in light-water reactors (LWRs). In the last few decades, many types of research studies have been conducted on these two phenomena separately, but the joint effect of these two mechanisms on the same crack has not been discussed yet though these two loads exist simultaneously in the LWRs. SCC is mainly a combination of the loading, the corrosive medium, and the susceptibility of materials while the LCF depends upon the elements such as compression, moisture, contact, and weld. As it is an attempt to combine SCC and LCF, this research focuses on the joint effect of SCC and LCF loading on crack propagation. The simulations are carried out using extended finite element method (XFEM) separately, for the SCC and LCF, on an identical crack. In the case of SCC, da/dt(mm/sec) is converted into da/dNScc (mm/cycle), and results are combined at the end. It has been observed that the separately calculated results for SCC da/dNScc and LCF da/dNm of crack growth rate are different from those of joint/overall effect,  da/dNom. By applying different SCC loads, the overall crack growth is measured as SCC load becomes the main cause of failure in LWRs in some cases particularly in the presence of residual stresses.

The Need for Strategic Planning for Nuclear Power Plant I&C Upgrades

2002 ◽  
Author(s):  
J. Douglas Hill ◽  
Paul Moore
Keyword(s):  
Strategic Planning ◽  
Control Systems ◽  
Digital Technology ◽  
Nuclear Power ◽  
Power Plants ◽  
Short Term ◽  
Nuclear Plants ◽  
And Control ◽  
Plant Control

Nuclear power plants rely on Instrumentation and Control (I&C) systems for control, monitoring and protection of the plant. The original, analog designs used in most nuclear plants have become or soon will be obsolete, forcing plants to turn to digital technology. Many factors affect the design of replacement equipment, including long-term and short-term economics, regulatory issues, and the way the plant operates on a day-to-day basis. The first step to all modernization projects should involve strategic planning, to ensure that the overall long and short-term goals of the plant are met. Strategic planning starts with a thorough evaluation of the existing plant control systems, the available options, and the benefits and consequences of these options.

Methodology to Evaluate Nuclear Power Plant Vulnerability to Events in the Transmission Grid

2014 ◽  
Author(s):  
Alberto Del Rosso ◽  
Jean-François Roy ◽  
Frank Rahn ◽  
Alejandro Capara
Keyword(s):  
Risk Assessment ◽  
Nuclear Power ◽  
Power Plants ◽  
Assessment Tool ◽  
Transmission Grid ◽  
Simulation Tools ◽  
Nuclear Plants ◽  
Western Electricity Coordinating Council ◽  
The Impact ◽  
The U.S

This paper presents a general approach to evaluate the risk of trip or Loss of Off-site Power (LOOP) events in nuclear power plants due to contingencies in the power grid. The proposed methodology is based on the Zone of Vulnerability concept for nuclear plants introduced by EPRI in previous work. The proposed methodology is intended to be part of an integrated probabilistic risk assessment tool that is being developed under ongoing EPRI R&D programs. A detailed analysis of many events occurred in actual nuclear plants has been performed in order to identify, classify and characterize the various vulnerability and type of failures that may affect a nuclear plant. Based the outcome of that analysis, a methodology for evaluating the impact of off-site transmission system events on nuclear plants has been outlined. It includes description of the type of contingencies and conditions that need to be included in the analysis, as well as provisions regarding the simulation tools and models that should be used in each case. The methodology is illustrated in a simplified representation of the Western Electricity Coordinating Council (WECC) system in the U.S.

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.

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.

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