• 한국압력기기공학회 논문집
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• 제18권2호
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• pp.37-42
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• 2022

An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.

• 대한기계학회논문집
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• 제19권4호
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• pp.1051-1061
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• 1995

In order perform leak-before-break design of nuclear piping systems and integrity evaluation of reactor vessels, full stress-strain (.sigma. - .epsilon.) curves and fracture resistance (J-R) curves are required. However it is time-consuming and expensive to obtain J-R curves experimentally. The objective of this paper is to develop two methods for J-R curve prediction. In the first method, elastic-plastic finite element analyses for a series of crack length / specimen width ratio were performed. Accordingly the load versus load line displacement (P .delta.) curve corresponding to the fracture strain is obtained and the J-R curve based on the generalized locus method is obtained. In the second method, the correlation between .sigma.-.epsilon. curves and J-R curves was statistically analyzed and an empirical equation to predict the J-R curve from the .sigma.-.epsilon. test result is proposed. A good correlation between the predicted results based on the proposed methods and the experimental ones is obtained.

• 대한기계학회논문집A
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• 제21권11호
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• pp.1796-1808
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• 1997

In order to perform leak-before-break design of nuclear piping systems and integrity evaluation of reactor vessels, full stress-strain curves and fracture resistance(J-R) curves are required. However it is time-consuming and expensive to obtain J-R curves experimentally. To resolve these problems, three different methods for predicting J-R curves from tensile data were proposed by the authors previously. The objective of this paper is to develop a computer program based on those J-R curve prediction methods. The program consists of two major parts ; the main program part for the J-R curve prediction and the database part. Several case studies were performed to verify the program, and it was shown that the predicted results were, in general, in good agreement with the experimental ones.

• 대한기계학회논문집A
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• 제27권1호
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• pp.8-17
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• 2003

Fracture behaviors of pipes with local wall thinning are very important for the integrity of nuclear Power Plant. In Pipes of energy Plants, sometimes, the local wall thinning may result from severe erosion-corrosion (E/C) damage. However, the effects of local wall thinning on strength and fracture behaviors of piping system were not well studied. In this paper, the monotonic bending tests were performed of full-scale carbon steel pipes with local wall thinning. A monotonic bending load was applied to straight pipe specimens by four-point loading at ambient temperature without internal pressure. From the tests, fracture behaviors and fracture strength of locally thinned pipe were manifested systematically. The observed failure modes were divided into four types; ovalization. crack initiation/growth after ovalization, local buckling and crack initiation/growth after local buckling. Also, the strength and the allowable limit of piping system with local wall thinning were evaluated.

• 한국압력기기공학회 논문집
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• 제10권1호
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• pp.107-112
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• 2014

The Code of Federal Regulations, Title 10, Part 50, Appendix H requires surveillance program for reactor pressure vessel(RPV) that the peak neutron fluence at the end of the design life of the vessel will exceed $1.0E+17n/cm^2$ (E>1.0MeV). 2D/1D Synthesis method based on DORT 3.1 transport calculation code has been widely used to determine fast neutron(E>1.0MeV) fluence exposure to RPV in the beltline region. RAPTOR-M3G(RApid Parallel Transport Of Radiation-Multiple 3D Geometries) performing full 3D transport calculation was developed by Westinghouse and KRIST(Korea Reactor Integrity Surveillance Technology) and applied for the evaluations of In-Vessel and Ex-Vessel neutron dosimetry. The reaction rates from measurement and calculation were compared and the results show good agreements each other.

• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.40-47
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• 2013

The capacity factor of nuclear power plant in Korea is the highest level in the world. However, the integrity assessment of nuclear power plant is depended on foreign country. Especially, most eddy current testing systems for inspecting steam generators in nuclear power plant are currently imported from USA, Canada, and so on. Therefore, the eddy current testing system can react more active and adaptive from economic and managerial standpoint for actual nuclear power plants in Korea is required. In this paper, an eddy current testing system for inspecting steam generators in nuclear power plants is introduced. Frequency generator, analog circuit, analog digital converter circuit, and digital control circuit are composed in eddy current testing system. A benchmarking of acquisition system and acquisition software, eddynet 11i made by Zetec, and modifications are carried out based on the test environment of Korea nuclear power plants. Finally, all eddy current apparatus are integrated to inspect steam generator tubes in nuclear power plants.

• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.20-24
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• 2013

Reactor Vessel Internals(RVIs), which are installed within the reactor pressure vessel and support the fuel assembly, take responsibility for safety of reactor core. In operating Nuclear Power Plants(NPPs), the RVIs have been exposed to severe conditions such as neutron irradiation, high temperature, high pressure, and high velocity of coolant flow and have degraded by materials aging with long-term operation. Therefore, the effective aging management plan and the appropriate regulatory requirements are necessary to maintain the integrity of RVIs. The purpose of this paper is to provide a review guide for Periodic Safety Review(PSR) of RVIs in presurized water reactor. The review guide is developed based on the revised review guides and reports established from IAEA and USNRC, and the analysis results of design characteristics, aging mechanisms, and operating experiences of RVIs in domestic and international NPPs. Consequently, the developed review guide for PSR of RVIs is expected to contribute an overall strategy and standard for the PSR of RVIs.

• 한국압력기기공학회 논문집
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• 제12권2호
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• pp.50-57
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• 2016

The study presents structural integrity assessment of ductility degradation of a baffle former assembly by performing finite element analysis considering real loading conditions and stress triaxiality. Variations of fracture strain curves of type 304 austenitic stainless steel with stress triaxiality are derived based on the previous study results. Temperature distributions during normal operation such as heat-up, steady state, and cool-down are calculated via finite element temperature analysis considering gamma heating and heat convection with reactor coolant. Variations of stress and strain state during long operation period are also calculated by performing sequentially coupled temperature-stress analysis. Fracture strain is derived by using the fracture curve and the stress triaxility. Finally, variations of ductility degradation damage indicator with the fracture strain and the equivalent inelastic strain are investigated. It is found that maximum value of the ductility degradation damage index continuously increases and becomes 0.4877 at 40 EFPYs. Also, the maximum value occurs at top and middle inner parts of the baffle former assembly before and after 20 EFPYs, respectively.

• 한국압력기기공학회 논문집
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• 제12권2호
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• pp.1-24
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• 2016

The original fracture criteria developed by Maxey/Kiefner for axial through-wall and surface-cracked pipes have worked well for many industries for a large variety of relatively low strength and toughness materials. However, newer line pipe steels have some unusual characteristics that differ from these older materials. One example is a test data that has demonstrated that X80 line-pipe with an axial through-wall-crack can fail at pressures about 30 percent lower than predicted with commonly used analysis methods for older steels. Thus, it is essential to review the currently available models and investigate the applicability of these models to newer high-strength line pipe materials. In this paper, the available models for predicting the failure behavior of axial-cracked pipes (through-wall-cracked and external surface-cracked pipes) were reviewed. Furthermore, the applicability of these models to high-strength steel pipes was investigated by analyzing limited full-scale pipe fracture initiation test results. Based on the analyzed results, the shortcomings of the available models were identified. For both through-wall and surface cracks, the major shortcomings were related to the characterization of the material toughness, which generally leads to non-conservative predictions in the J-T analyses. The findings in this paper may be limited to the test data that were consider for this study. The requisite characteristics of a potential model were also identified in the present paper.

• 한국압력기기공학회 논문집
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• 제14권1호
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• pp.8-14
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• 2018

Printed circuit heat exchangers (PCHEs) are widely used with an increasing demand for industrial applications. PCHEs are capable of operating at high temperatures and pressure. We consider a PCHE as a candidate intermediate heat exchanger type for a high temperature gas-cooled reactor (HTGR). For conventional application using stainless steels, design and manufacturing of PCHEs are well established. For applications to HTGR, knowledge of longitudinal conduction and deformation of channel is required to estimate design margin. This paper analyzes the effects of longitudinal conduction and deformation of channel on thermal performance using a code internally developed for design and analysis of PCHEs. The code has a capability of two dimensional simulations. Longitudinal conduction is estimated using the code. In HTGR operating condition, about ten percent of design margin is required to compensate thermal performance. The cross-sectional images of PCHE channels are obtained using an optical microscope. The images are processed with computer image process technique. We quantify the deformation of channel with dimensional parameters. It is found that the deformation has negative effect on structural integrity. The deformation enhances thermal performance when the shape of channel is straight in laminar flow regime. It reduces thermal performance in cases of a zigzag channel and turbulent flow regime.