• 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.

• 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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.6
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• pp.637-647
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• 2014

In pressurized water reactors, the upper head of the reactor pressure vessel (RPV) contains numerous control rod drive mechanism (CRDM) nozzles. These nozzles are fabricated by welding after being inserted into the RPV head with a room temperature shrink fit. The tensile residual stresses caused by this welding are a major factor in primary water stress corrosion cracking (PWSCC). Over the last 15 years, the incidences of cracking in alloy 600 CRDM nozzles have increased significantly. These cracks are caused by PWSCC and have been shown to be driven by the welding residual stresses and operational stresses in the weld region. Various measures are being sought to overcome these problems. The defects resulting from the welding process are often the cause of PWSCC acceleration. Therefore, any weld defects found in the RPV manufacturing process are immediately repaired by repair welding. Detailed finite-element simulations for the Korea Nuclear Reactor Pressure Vessel were conducted in order to predict the magnitudes of the repair weld residual stresses in the tube materials.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.817-824
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• 2014

To ensure the safety of research reactors, the water level must be maintained above the required height. When a pipe ruptures, the siphon phenomenon causes continuous loss of coolant until the hydraulic head is removed. To protect the reactor core from this kind of accident, a siphon breaker has been suggested as a passive safety device. This study mainly focused on two variables: the size of the pipe rupture and the timing of air entrainment. In this study, the size of the pipe rupture was increased to the guillotine break case. There was a region in which a larger pipe rupture did not need a larger siphon breaker, and the water flow rate was related to the size of the pipe rupture and affected the residual water quantity. The timing of air entrainment was predicted to influence residual water level. However, the residual water level was not affected by the timing of air entrainment. The experimental cases, which showed the characteristic of partical sweep-out mode in the separation of siphon breaking phenomenon [2], showed almost same trend of physical properties.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1547-1554
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• 2017

VESTA (Verification of Ex-vessel corium STAbilization) and VESTA-S (-small) test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging $ZrO_2$ melt jet on a sacrificial material were performed to investigate the ablation characteristics. $ZrO_2$ melt in an amount of 65-70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40), and the other is a stainless steel (SUS304) melt. Metallic melt in an amount of 1.5-2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. $ZrO_2$ melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is $UO_2$ 60%, Zr 10%, $ZrO_2$ 15%, SUS304 14%, and $B_4C$ 1%, was melted in a cold crucible using an induction heating technique.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.10
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• pp.1337-1345
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• 2011

In pressurized water reactors (PWRs), the reactor pressure vessel (RPV) upper head contains numerous control rod drive mechanism (CRDM) nozzles. In the last 10 years, the incidences of cracking in alloy 600 CRDM nozzles and their associated welds has increased significantly. Several axial and circumferential cracks have been found in CRDM nozzles in European PWRs and U.S. nuclear power plants. These cracks are caused by primary water stress corrosion cracking (PWSCC) and have been shown to be driven by welding residual stresses and operational stresses in the weld region. Therefore, detailed finite-element (FE) simulations for the Korea Nuclear Reactor Pressure Vessel have been conducted in order to predict the magnitudes of the weld residual stresses in the tube materials. In particular, the weld residual stress results are compared in terms for nozzle location, geometry factor$r_o$/t, geometry of fillet, and adjacent nozzle.

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.3
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• pp.256-263
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• 2009

After several PWSCCs were found in Bugey(France), Ringhals(Sweden), Tihange(Belgium), Oconee, Arkansas, Crystal Fever, Davis-Basse, VC Summer(U.S.A.), Thuruga(Japan), USNRC and PNNL started the research on PWSCC, that is, the PINC project. USNRC required KINS to participate in the PINC project in May 2005. KINS organized the Korean consortium at March 2006 and Pre-RRT for RVHP were performed for the preparation of PINC RRT. Through these preliminary RRT, Korea NDE teams can learn and develop the detection and sizing technique for RVHP dissimilar metal weld. These techniques are now being prepared in Korea and need to be utilized for the In-service inspection of the RVHP and BMI of Korea Nuclear Power Plants. PINC RRT mock-ups will be helpful to training.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.10
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• pp.1115-1130
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• 2012

In nuclear power plants, the reactor pressure vessel (RPV) upper head control rod drive mechanism (CRDM) penetration nozzles are fabricated using J-groove weld geometry. Recently, the incidences of cracking in Alloy 600 CRDM nozzles and their associated welds have increased significantly. The cracking mechanism has been attributed to primary water stress corrosion cracking (PWSCC), and it has been shown to be driven by welding residual stresses and operational stresses in the weld region. The weld-induced residual stress is the main factor contributing to crack growth. Therefore, an exact estimation of the residual stress is important for ensuring reliable operation. This study presents the residual stress computation performed for an RPV CRDM penetration nozzle in Korea. Based on two and three dimensional finite element analyses, the effect of welding variables on the residual stress variation is estimated for sensitivity analysis.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.1
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• pp.1-5
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• 2010

Acoustic emission(AE) has emerged as a powerful nondestructive tool to detect or monitor preexisting defects and leaks in the vessel structures. A Bragg grating based acoustic emission sensor system is developed. Various type of fiber Bragg grating sensor including the variable length of sensing part was fabricated and prototype sensor system was tested by using PZT pulser and pencil lead break sources. Two types of sensor attachment were used. First, the fiber Bragg grating sensor was attached fully to the surface using bonding agent. Second one is that one part of fiber was attached to the surface partly by bonding and the other part of fiber will be act as a cantilever. That is, the resonant frequency of the fiber Bragg grating sensor will depend on the length of sensing part. The final goal of the sensor system is to provide on-line monitoring of cracks or leaks in reactor vessel head penetration of nuclear power plants.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.9
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• pp.1047-1052
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• 2012

Primary water stress corrosion cracking (PWSCC) instances have been reported in the Alloy 600 reactor pressure vessel head penetration nozzle and the Alloy 82/182 dissimilar metal butt weld nozzle in several PWRs. Therefore, in-service inspection programs have been adopted worldwide to prevent failure at the weld region. If a PWSCC is observed at the dissimilar metal weld region during inspection, its structural integrity should be evaluated; however, this requires considerable time and effort, and this might lead to a decrease in the plant utilization coefficient. To prevent this, KHNP-CRI have established integrity assessment criteria and developed a computer program for the fast evaluation and judgment of PWSCC. In this paper, the results and current status of the same are presented. Through this study, criteria for the structural integrity evaluation of PWSCC have been established, and a computer program has been developed to realize technical means for the evaluation of PWSCC structural integrity.