• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1423-1432
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• 2016

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the "screening criterion" for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no "screening criterion". In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

• Korean Journal of Metals and Materials
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• v.48 no.2
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• pp.122-132
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• 2010

Higher strength and fracture toughness of reactor pressure vessel steels can be obtained by changing the material specification from that of Mn-Mo-Ni low alloy steel (SA508 Gr.3) to Ni-Mo-Cr low alloy steel (SA508 Gr.4N). However, the operation temperature of the reactor pressure vessel is more than $300^{\circ}C$ and the reactor operates for over 40 years. Therefore, we need to have phase stability in the high temperature range in order to apply the SA508 Gr.4N low alloy steel for a reactor pressure vessel. It is very important to evaluate the temper embrittlement phenomena of SA508 Gr.4N for an RPV application. In this study, we have performed a Charpy impact test and tensile test of SA508 Gr.4N low alloy steel with changing impurity element contents such as Mn and P. And also, the mechanical properties of these low alloy steels after longterm heat treatment ($450^{\circ}C$, 2000hr) are evaluated. Further, evaluation of the temper embrittlement by fracture analysis was carried out. Temper embrittlement occurs in KL4-Ref and KL4-P, which show a decrease of the elongation and a shifting of the transition curve toward high temperature. The reason for the temper embrittlement is the grain boundary segregation of the impurity element P and the alloying element Ni. However, KL4-Ref shows temper embrittlement phenomena despite the same contents of P and Ni compared with SC-KL4. This result may be caused by the Mn contents. In addition, the behavior of embrittlement is not largely affected by the formation of $M_3P$ phosphide or the coarsening of Cr carbides.

• Korean Journal of Metals and Materials
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• v.47 no.9
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• pp.533-541
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• 2009

Materials used for a reactor pressure vessel(RPV) are required high strength and toughness, which determine the safety margin and life of a reactor. Ni-Mo-Cr low alloy steel shows better mechanical properties than existing RPV steels due to higher Ni and Cr contents compared to the existing RPV steels. The present study focuses on effects of Ni, Cr contents on the cleavage fracture toughness of Ni-Mo-Cr low alloy steels in the transition temperature region. The fracture toughness was characterized by a 3-point bend test of precracked Charpy V-notch(PCVN) specimens based on ASTM E1921-08. The test results indicated that the fracture toughness was considerably improved with an increase of Ni and Cr contents. Especially, control of Cr content was more effective in improving fracture toughness than manipulating Ni content, though Charpy impact toughness was changed more extensively by adjusting Ni content. These differences between changes in the fracture toughness and that in the impact toughness were derived from microstructural features, such as martensite lath size and carbide precipitation behavior.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2002-2010
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• 2024

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

• 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 Radiation Protection and Research
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• v.28 no.2
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• pp.109-116
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• 2003

Based on the necessity to evaluate the activation products inventory during decommissioning lot domestic nuclear power plants, a preliminary estimation of the activation products inventory for Kori unit 1, which is getting close to the end of lifetime, was carried out with ANISN and ORIGEN2 code. In order to calculate neutron nux using ANISN code, the reactor was divided into 9 zones from core to bioshield concrete for radial direction. Also :he cross-section of main nuclides were calibrated with neutron flux in the reactor pressure vessel(RPV) region. The results showed that 95 % of tile total radioactivity in RPV from reactor shutdown to 10 years came from the nuclides of $^{55}Fe,\;^{59}Ni,\;^{63}Ni\;and\;^{60}Co$. And the total radioactivity with cooling of more than 50 years after decommissioning was no more than 0.2 % of at the time of shutdown. Considering the weight of RPV is 210 tons, the total radioactivity of RPV reached to $5.25{\times}10^{6}GBq$ at shutdown time. As compared with the total radioactivity of bioshield concrete at reactor shutdown time, the radioactivity after tooling more than 10 years was below 1 %.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.210-210
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• 1998

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.613-614
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• 2005

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.3
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• pp.505-513
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• 2002

During heatup and cooldown of pressurized water reactor, thermal stress was generated in the reactor pressure vessel (RPV) because of the temperature gradient. To prevent potential failure of RPV, pressure was required to be maintained below the P-T limit curves. In this paper, several methods for constructing the P-T limit curves including the ASME Sec. XI, App. G method were explained and the results were compared. Then, the effects of the various parameters such as flaw size, flaw orientation, cooldown rate, existence of chad, and reference fracture toughness, were evaluated. It was found that the current ASME Sec. XI App. G method resulted in the most conservative P-T limit curve. As the more accurate fracture mechanics analysis results were used, some of the conservatism can be removed. Among the parameters analysed, reference flaw orientation and reference fracture toughness curve had the greatest effect on the resulting P-T limit curves.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.11
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• pp.2399-2408
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• 2002

A prediction procedure has been developed to evaluate the microtructures and material properties of heat affected zone (HAZ) in pressure vessel steel weld, based on temperature analysis, thermodynamics calculation and reaction kinetics model. Temperature distributions in HAE are calculated by finite element method. The microstructures in HAZ are predicted by combining the temperature analysis results with the reaction kinetics model for austenite grain growth and austenite decomposition. Substituting the microstructure prediction results into the previous experimental relations, the mechanical material properties such as hardness, yielding strength and tensile strength are calculated. The prediction procedure is modified and verified by the comparison between the present results and the previous study results for the simulated HAZ in reactor pressure vessel (RPV) circurnferential weld. Finally, the microstructures and mechanical material properties are determined by applying the final procedure to real RPV circumferential weld and the local weak zone in HAZ is evaluated based on the application results.