• 한국소음진동공학회논문집
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• 제12권11호
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• pp.856-863
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• 2002

Seismic qualification of the main control board(MCB) for the nuclear power plant Ulchin 5 and 6 has been performed with the guideline of ASME Section III and IEEE 344 code. As the size and weight of the MCB are too large and heavy to excite using the excitation table, finite element analysis is used in order to investigate the dynamic behaviors and structural integrity of the MCB. As the fundamental frequencies of the equipment are found to be less than 33 Hz, which is the upper frequency limit for the dynamic analysis, response spectrum analysis using ANSYS is performed in order to combine the modal stresses within the frequency limit. In order to confirm the electrical stability of the major components of the MCB. modal analysis theory has been adopted to derive the required response spectra at the component locations. As the all combined stresses obtained from the above procedures are less than the allowable stresses and no mechanical or electrical failures are found from the seismic testing, the authors can confirm the safety of the nuclear equipment MCB under the given seismic loading conditions.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1681-1688
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• 2019

In accordance with ASME Boiler and Pressure Vessel (B&PV) Code Sec.XI Appendix. G, a postulated crack is located at the beltline of a reactor pressure vessel because the neutron flux at the beltline is higher than elsewhere. This means that the distance between the core and the semi-spherical bottom head is longer than the distance between the core and the cylindrical beltline. However, several Small and Medium sized Reactors have bottom heads with diverse shapes, including dished or semi-elliptical shapes, to satisfy the requirement and performance. So, the aim of this paper is to evaluate the effect of crack location on Pressure-Temperature limit curve. To do this, two types of postulated crack location, such as beltline and semi-elliptical bottom head, were adopted to derive the Pressure-Temperature limit curve. Also, parametric studies for neutron flux, crack shape and so on were performed. As a result, core critical temperature of semi-elliptical bottom head is found to higher than that of beltline even when they have same values of thickness and neutron flux. This result will be useful to enhance the understanding of Pressure-Temperature limit curve.

• International Journal of Naval Architecture and Ocean Engineering
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• 제12권1호
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• pp.868-880
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• 2020

One of the main concerns in the structural integrity of offshore pipelines is mechanical damage from external loads. Pipelines are exposed to fatigue failure in welded joints due to geometric discontinuity. In addition, fatigue loads such as currents, waves, and platform motions may cause significant plastic deformation and fracture or leakage within a relatively low-cycle regime. The 2007 ASME Div. 2 Code adopts the master S―N curve for the fatigue evaluation of welded joints based on the mesh-insensitive structural stress. An extension to the master S―N curve was introduced to evaluate the low-cycle fatigue strength. This structural strain method uses the tensile properties of the material. However, the monotonic tensile properties have limitations in describing the material behavior above the elastic range because most engineering materials exhibit hardening or softening behavior under cyclic loads. The goal of this study is to extend the cyclic stress-strain behavior to the structural strain method. To this end, structural strain-based procedure was established while considering the cyclic stress-strain behavior and compared to the structural strain method with monotonic tensile properties. Finally, the improved prediction method was validated using fatigue test data from full-scale girth-welded pipes.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2188-2197
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• 2022

To ensure the safety margin of a reactor pressure vessel (RPV) under normal operating conditions, it is regulated through the pressure-temperature (P-T) limit curve. The stress intensity factor (SIF) obtained by the internal pressure and thermal load should be obtained through crack analysis of the nozzle corner crack in advance to generate the P-T limit curve for the nozzle. In the ASME code Section XI, Appendix G, the SIF via the internal pressure for the nozzle corner crack is expressed as a function of the cooling or heating rate, and the wall thickness, however, the SIF via the thermal load is presented as a polynomial format based on the stress linearization analysis results. Inevitably, the SIF can only be obtained through finite element (FE) analysis. In this paper, simple prediction equations of the SIF via the thermal load under, cool-down and heat-up conditions are presented. For the Korean standard nuclear power plant, three geometric variables were set and 72 cases of RPV models were made, and then the heat transfer analysis and thermal stress analysis were performed sequentially. Based on the FE results, simple engineering solutions predicting the value of thermal SIF under cool-down and heat-up conditions are suggested.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2386-2394
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• 2022

In this paper, smooth and notched bar tensile tests of austenitic stainless steel 304 are performed, covering four different multi-axial stress states and six different strain rate conditions, to investigate the effect of the stress triaxiality and strain rate on fracture strain. Test data show that the measured true fracture strain tends to decrease with increasing stress triaxiality and strain rate. The test data are then quantified using the Johnson-Cook (J-C) fracture strain model incorporating combined effects of the stress triaxiality and strain rate. The determined J-C model can predict true fracture strain overall conservatively with the difference less than 20%. The conservatism in the strain-based acceptance criteria in ASME B&PV Code, Section III, Appendix FF is also discussed.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.1085-1097
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• 2022

A printed circuit steam generator (PCSG) is being considered as the component for pressurized water reactor (PWR) type small modular reactor (SMR) that can further reduce the physical size of the system. Since a steam generator in many PWR-type SMR generates superheated steam, it is expected that dryout front oscillation can potentially cause thermal fatigue failure due to cyclic thermal stresses induced by the transition in boiling regimes between convective evaporation and film boiling. To investigate the fatigue issue of a PCSG, a reference PCSG is designed in this study first using an in-house PCSG design tool. For the stress analysis, a finite element method analysis model is developed to obtain the temperature and stress fields of the designed PCSG. Fatigue estimation is performed based on ASME Boiler and pressure vessel code to identify the major parameters influencing the fatigue life time originating from the dryout front oscillation. As a result of this study, the limit on the temperature difference between the hot side and cold side fluids is obtained. Moreover, it is found that the heat transfer coefficient of convective evaporation and film boiling regimes play an essential role in the fatigue life cycle as well as the temperature difference.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1726-1746
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• 2022

This paper presents fracture mechanics analysis results for various cracks located at pressurized water reactor pressure vessel nozzle crotch corners taking into consideration constraint effect. Technical documents such as the ASME B&PV Code, Sec.XI were reviewed and then a fracture mechanics analysis procedure was proposed for structural integrity assessment of various nozzle crotch corner cracks under normal operation conditions considering the constraint effect. Linear elastic fracture mechanics analysis was performed by conducting finite element analysis with the proposed analysis procedure. Based on the evaluation results, elastic-plastic fracture mechanics analysis taking into account the constraint effect was performed only for the axial surface crack of the reactor pressure vessel outlet nozzle with cladding. The fracture mechanics analysis result shows that only the axial surface crack in the reactor pressure vessel outlet nozzle has the stress intensity factor exceeding the low bound of upper-shelf fracture toughness irrespectively of considering the constraint effect. It is confirmed that the J-integral for the axial crack of the outlet nozzle does not exceed the ductile crack initiation toughness. Hence, it can be ensured that the structural integrity of all the cracks is maintained during the normal operation.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1580-1586
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• 2021

The application of induction bending processes to industrial pipe production is increasing. The induction bending process has the effect of reducing the number of inspections and preventing leaks by reducing the weld of the pipe. For these reasons, efforts have been made to apply an induction bending process to the pipe of the PGSFR under development in Korea and this is the first attempt in the SFR design. Since the PGSFR pipe has a relatively large diameter-to-thickness ratio, it is difficult to fabricate an induction bending pipe that meets the requirements. In addition, the material properties may change because the pipe heats to a very high temperature during the induction bending process. In this study, P91 pipes were fabricated by induction bending, and the results from analyzing the induction bending process' applicability to the P91 pipe of the PGSFR are examined. The various dimensional measurements of the pipes fabricated by the induction bending process were surveyed to determine whether the requirements of the ASME Code were met. The minimum thickness, ovality, and wall buckling measured in the fabricated pipe met all the requirements. Tensile, impact, and hardness tests at various locations of the fabricated pipe also satisfied the requirements.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3571-3580
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• 2023

The application of the induction bending process to pipe systems in various industrial fields is increasing. Recently, efforts have also been made to apply this bending process to nuclear power plants because it can innovatively reduce welded parts of the curved pipes, such as elbows. However, there have been no cases of the application of induction bending to the piping of nuclear power plants. In this study, the applicability of the P91 induction bending piping for the sodium-cooled fast reactor PGSFR was validated through high temperature low cycle fatigue tests and creep tests using P91 induction bending pipe specimens. The tests confirmed that the materials sufficiently satisfied the fatigue life and the creep rupture life requirements for P91 steel at 550 ℃ in the ASME B&PV Code, Sec. III, Div. 5. The results show that the effects of heating and bending by the induction bending process on the material properties were not significant and the induction bending process could be applicable to piping system of PGSFR well.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4477-4490
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• 2023

Fuel locking mechanisms (FLMs) are essential in upward-flow research reactors to prevent accidental fuel separation from the core during reactor operation. This study presents a novel design concept for a remotely controlled plate-type nuclear fuel locking mechanism. By employing electromagnetic field analysis, we optimized the design of the electromagnet for fuel unlocking, allowing the FLM to adapt to various research reactor core designs, minimizing installation space, and reducing maintenance efforts. Computational flow analysis quantified the drag acting on the fuel assembly caused by coolant upflow. Subsequently, we performed finite element analysis and evaluated the structural integrity of the FLM based on the ASME boiler and pressure vessel (B&PV) code, considering design loads such as dead weight and flow drag. Our findings confirm that the new FLM design provides sufficient margins to withstand the specified loads. We fabricated a prototype comprising the driving part, a simplified moving part, and a dummy fuel assembly. Through basic operational tests on the assembled components, we verified that the manufactured products meet the performance requirements. This remote-controlled micro locking mechanism holds promise in enhancing the safety and efficiency of plate-type nuclear fuel operation in upflow research reactors.