• Nuclear Engineering and Technology
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• 제44권5호
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• pp.543-550
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• 2012

The decrease in the fracture toughness of ferritic steels in a reactor pressure vessel is an important factor in determining the lifetime of a nuclear power plant. A surveillance program has been in place in Korea since 1979 to assess the structural integrity of RPV steels. In this work, the surveillance data were collected and analyzed statistically in order to derive the empirical relationship between the embrittlement and strengthening of irradiated reactor pressure vessel steels. There was a linear relationship between the yield strength change and the transition temperature shift change at 41 J due to irradiation. The proportional coefficient was about $0.5^{\circ}C$/MPa in the base metals (plate/forgings). The upper shelf energy decrease ratio was non-linearly proportional to the yield strength change, and most of the data lay along the trend curve of the US results. The transition regime temperature interval, ${\Delta}T_T$, was less than the US data. The overall change from irradiation was very similar to the US results. It is expected that the results of this study will be applied to basic research on the multiscale modeling of the irradiation embrittlement of RPV materials in Korea.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(3)
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• pp.290-295
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• 1996

The purpose of this study is to establish the pressure/temperature curves of Reactor Coolant System for brittle fracture prevention. The pressure/temperature curve is the basis to select RC Pump and limits to operate the plant. Based on the plant operation experience, this curve should be re-generated periodically in order to ensure the structural integrity using data from the test of reactor vessel surveilance materials to compensate for the irradiation effects. This study provides the procedure of pressure/temperature curve generation in term of brittle fracture prevention of reactor vessel. Using the UCN 3&4 data, the sample pressure/temperature curve was generated, and it was compared with those of YGN 3&4 based on the stress and $RT_{NDT}$value.

• Nuclear Engineering and Technology
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• 제44권6호
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• pp.683-688
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• 2012

The structural integrity of mechanical components during several transients should be assured in the design stage. This requires a fatigue analysis including thermal and stress analyses. As an example, this study performs a fatigue analysis of the reactor pressure vessel of SMART during arbitrary transients. Using heat transfer coefficients determined based on the operating environments, a transient thermal analysis is performed and the results are applied to a finite element model along with the pressure to calculate the stresses. The total stress intensity range and cumulative fatigue usage factor are investigated to determine the adequacy of the design.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3229-3240
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• 2023

Due to neutron radiation, atomic displacement has a significant effect on material in nuclear reactors. A range of secondary displacement models, including the Kinchin-Pease (K-P), Lindhard, Norgett-Robinson-Torrens (NRT), and athermal recombination-corrected displacement per atom (arc-dpa) have been suggested to calculate the number of displacement per atom (dpa). As neutron elastic interaction is the main cause of displacement damage, the focus of the current study is to calculate the atomic displacement caused by the neutron elastic interaction in order to estimate the exact amount of yield strength in a WWER-1000 reactor pressure vessel. To achieve this purpose, the reactor core is simulated by MCNPX code. In addition, a program is developed to calculate the elastic radiation damage induced by the incident neutron flux (RADIX) based on different models using Fortran programming language. Also, due to non-elastic interaction, the displacement damage is calculated by the HEATR module of the NJOY code. ASME E-693-01 standard, SPECTER, NJOY codes, and other pervious findings have been used to validate RADIX results. The results showed that the RADIX(arc-dpa)/HEATR outputs have appropriate accuracy. The relative error of the calculated dpa resulting from RADIX(arc-dpa)/HEATR is about 8% and 46% less than NJOY code, respectively in the ¼ and ¾ vessel wall.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.545-553
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• 2016

The failure probabilities of the reactor pressure vessel (RPV) for low temperature over-pressurization (LTOP) and cool-down transients are calculated in this study. For the cool-down transient, a pressure-temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME) code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition ($RT_{NDT}$). The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집A
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• pp.782-787
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• 2001

This paper is concerned with the global rupture of a nuclear reactor pressure vessel(RPV) in a severe accident. During the severe reactor accident of molten core, the temperature and the pressure in the nuclear reactor rise to a certain level depending on the initial and subsequent condition of a severe accident. While the rise of the temperature cause the thermal softening of RPV material, the rise of the internal pressure could cause failure of the RPV lower head. The global rupture of an RPV is simulated by finite element limit analysis for the collapse pressure and mode and this analysis results have been compared with a variation of the internal pressure of RPV. The finite element limit method is a systematic tool to secure the safety criteria of a nuclear reactor and to evaluate the in-vessel corium retention.

• 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.

• Structural Engineering and Mechanics
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• 제37권3호
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• pp.257-265
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• 2011

Plant-specific analyses of 5 types of domestic reactors in Korea are performed to assure the structural integrity of the reactor pressure vessel (RPV) during transients which are expected to initiate pressurized thermal shock (PTS) events. The failure probability of the RPV due to PTS is obtained by performing probabilistic fracture mechanics analysis. The through-wall cracking frequency is calculated and compared to the acceptance criterion. Considering the fluence at the end of life expected by surveillance test, the sufficient safety margin is expected for the structural integrity of all reactor pressure vessels except for the oldest one during the pressurized thermal shock events. If the flaw with aspect ratio of 1/12 is considered to eliminate the conservatism, the acceptance criteria is not exceeded for all plants until the fluence level of $8{\times}10^{19}\;n/cm^2$, generating sufficient margin beyond the design life.

• 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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• 제54권1호
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• pp.401-413
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• 2022

This paper provides a simple method by which to estimate the cross-section stress profiles for nozzles designed according to ASME Code Section III. Further, this method validates the effectiveness of earlier work performed by the authors on standard nozzles. The method requires only the geometric information of the pressure vessel and the attached nozzle. A PWR direct vessel injection nozzle, a PWR outlet nozzle, a PWR inlet nozzle and a BWR recirculation outlet nozzle are selected based on their corresponding specific designs, e.g., a varying nozzle radius, a varying nozzle thickness and an outlet nozzle boss. A cross-section stress profile comparison shows that the estimates are in good agreement with the finite element analysis results. Differences in stress intensity factors calculated in accordance with ASME BPVC Section XI Appendix G are discussed. In addition, a change in the dimensions of an alternate nozzle design relative to the standard values is discussed, focusing on the stress concentration factors of the nozzle inside corner.