• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.44-50
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• 2014

There were many flaw issues of reactor vessel head penetration in USA fleets. USNRC issued 10CFR50.55a to implement reactor vessel head penetration ultrasonic examination performance demonstration(PD) in US for enhancement of inspection reliability. After September 2009, all US utilities inspected their RVHP with PD qualified system. Korea Hydro and Nuclear Power Company(KHNP) have developed reactor vessel head penetration performance demonstration system for ultrasonic test to apply for pressurized light-water reactor power plants in accordance with 10CFR50.55a since September 2011. RVHP configuration surveying and analysis, code requirement analysis, and performance demonstration specimen design were performed up to this day. Fingerprinting of manufactured specimen, development of test data management program, development of operation procedure, input of flawed data, and development of final report will be performed for the next step. This paper describes the development status of the performance demonstration system for reactor vessel head penetration ultrasonic examination in Korea.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.9-15
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• 2010

US NRC issued rulemaking of 10CFR50.55a to perform the Perservice and Inservice inspection for Reactor Vessel Head Penetration Nozzle of US Nuclaer plant. The rulemaking was required the EPRI Demonstration to verify the NDE technique performing special Ultrasonic examination. In order to meet this requirement, the UT and ECT procedures was demonstrated and the NDE personnel were qualified by EPRI. In this paper, the NDE technique and analysis method are described the Preservice inspection for the Palo Verde #1/2/3 Replacement Reactor Vessel Head Penetration Nozzle using the qualified procedures and personnel.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.27 no.3
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• pp.191-198
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• 2014

In this paper, structural integrity evaluation of reactor pressure vessel bottom head without penetration nozzles in core melting accident has been performed. Considering the analysis results of thermal load, weight of molten core debris and internal pressure, thermal load is the most significant factor in reactor vessel bottom head. The failure probability was evaluated according to the established failure criteria and the evaluation showed that the equivalent plastic strain results are lower than critical strain failure criteria. Thermal-structural coupled analyses show that the existence of elastic zone with a lower stress than yield strength is in the middle of bottom head thickness. As a result of analysis, the elastic zone became narrow and moved to the internal wall as the internal pressure increases, and it is evaluated that the structural integrity of reactor vessel is maintained under core melting accident.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1524-1536
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• 2017

During the fabrication process of reactor vessel head penetration (RVHP), the grain size of the tube material can be changed by hot or cold work and the inner side of the tube can also be shrunk due to welding outside of the tube. Several nonregular time-of-flight diffraction (TOFD) signals were found because of deformed grains. In this paper, an investigation of nonregular TOFD indications acquired from RVHP tubes using experiments and computer simulation was performed in order to identify and distinguish TOFD signals by coarse grains from those by Primary Water Stress Corrosion Crack (PWSCC). For proper understanding of the nonregular TOFD indications, microstructural analysis of the RVHP tubes and prediction of signals scattered from the grains using Finite Element Method (FEM) simulation were performed. Prediction of ultrasonic signals from the various sizes of side drilled holes to find equivalent flaws, determination of the size of the nonregular TOFD indications from the coarse grains, and experimental investigation of TOFD signals from coarse grain and shrinkage geometry to identify PWSCC signals were performed. From the computer simulation and experimental investigation results, it was possible to obtain the nonregular TOFD indications from the coarse grains in the alloy 690 penetration tube of RVHP; these nonregular indications may be classified as PWSCC. By comparing the computer simulation and experimental results, we were able to confirm a clear difference between the coarse grain signal and the PWSCC signal.

• 대한방사선방어학회:학술대회논문집
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• 2009.04a
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• pp.304-305
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• 2009

• Proceedings of the KWS Conference
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• 2005.06a
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• pp.185-187
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• 2005

The reactor pressure vessel head of PWR has penetrations for control rod drive mechanism and instrumentation systems. The Primary coolant water and operating temperature can cause the stress-corrosion cracking of these nickel-based alloy penetrations. It is difficult to detect and size flaws such as SCC in the reactor head penetrations using conventional W methods because of complex geometry, Therefore, the utilities are using the TOFD technique for the detection and sizing of the flaw. This study shows the correlation between the ultrasonic wave direction and the orientation of the flaw and the range of flaw depth which can be detected by the TOFD techniques.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.9
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• pp.1159-1168
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• 2013

In pressurized water nuclear reactors (PWRs), the reactor pressure vessel (RPV) upper head contains penetration nozzles that use a control rod drive mechanism (CRDM). The penetration nozzle uses J-groove weld geometry. Recently, the occurrence of cracking in alloy 600 CRDM penetration nozzle has increased. This is attributable to primary water stress corrosion cracking (PWSCC). PWSCC is known to be susceptible to the welding residual stress and operational stress. Generally, the tensile residual stress is the main factor contributing to crack growth. Therefore, this study investigates the effect on weld residual stress through different analysis methods for normal operating conditions using finite element analysis. In addition, this study also considers the effect of repeated normal operating condition cycles on the weld residual stress. Based on the analysis result, this paper presents a normal operating condition analysis method.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.56-61
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• 2013

This paper presents the results of integrity assessment for the cracks happened in reactor vessel upper head penetration nozzles. The crack morphology for a boat sample from crack area was analyzed through microscope. The stress condition including weld residual stress around crack was analyzed using finite element analysis. From the results of crack morphology and stress condition, the crack was concluded as primary water stress corrosion cracking. The integrity of the cracked nozzle was assessed by the methodology provided in ASME Section XI. According to the assessment results, the remaining life of the cracked nozzle was 1.43 yrs. and the plant decided to repair it.

• Proceedings of the Korean Society of Machine Tool Engineers Conference
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• 2005.05a
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• pp.216-227
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• 2005

The head penetrations for control rod drive mechanism and instrumentation systems are installed at the reactor pressure vessel head of PWRs. Primary coolant water and the operating conditions of PWR plants can cause cracking of these nickel-based alloy through a process called primary water stress corrosion cracking (PWSCC). Inspection of the head penetrations to ensure the integrity of the head penetrations has been interested since reactor coolant leakages were found at U. S. reactors in 2000 and 2001. The complex geometry of the head penetrations and the very low echo amplitude from the fine, multiple flaws due to the nature of the see made it difficult to detect and size the flaws using conventional pulse-echo UT methods. Time-of-flight-diffraction technique, which utilizes the time difference between the flaw tips while pulse-echo does the flaw response amplitude from the flaw, has been selected for this inspection for it's best performance of the detection and sizing of the head penetration see flaws. This study defines the limits of the detectable and accurately sizable minimum flaw size which can be detected by the General TOFD and the Delta TOFD techniques for circumferentially and axially oriented flaws respectively. These results assures the reliability of the inspection techniques to detect and accurately size for various kind of flaws, and will also be utilized for the future development and qualifications of the TOFD techniques to enhance the detecting sensitivity and sizing accuracy of the flaws of the reactor head penetrations in nuclear power plants.

• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.4
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• pp.273-280
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• 2016

The simulation of nondestructive testing has been used in the prediction of the signal characteristics of various defects and in the development of the procedures. CIVA, a simulation tool dedicated to nondestructive testing, has good accuracy and speed, and provides a three-dimensional graphical user interface for improved visualization and familiar data displays consistent with an NDE technique. Even though internal validations have been performed by the CIVA software development specialists, an independent validation study is necessary for the assessment of the accuracy of the software prior to practical use. In this study, time of flight diffraction signals of ultrasonic inspection of a calibration block for reactor vessel head penetrations were simulated using CIVA. The results were compared to the experimentally inspected signals. The accuracy of the simulated signals and the possible range for simulation were verified. It was found that, there is a good agreement between the CIVA simulated and experimental results in the A-scan signal, B-scan image, and measurement of depth.