• Nuclear Engineering and Technology
• /
• v.32 no.4
• /
• pp.410-423
• /
• 2000

A two-dimensional numerical model is developed and applied to the LAVA-EXV tests performed at the Korea Atomic Energy Research Institute (KAERI) to investigate the external cooling effect on the thermal margin to failure of a reactor pressure vessel (RPV) during a severe accident. The computational program was written to predict the temperature profile of a two-dimensional spherical vessel segment accounting for the conjugate heat transfer mechanisms of conduction through the debris and the vessel, natural convection within the molten debris pool, and the possible ablation of the vessel wall in contact with the high temperature melt. Results of the sensitivity analysis and comparison with the LAVA-EXV test data indicated that the developed computational tool carries a high potential for simulating the thermal behavior of the RPV during a core melt relocation accident. It is concluded that the main factors affecting the RPV failure are the natural convection within the debris pool and the ablation of the metal vessel, The simplistic natural convection model adopted in the computational program partly made up for the absence of the mechanistic momentum consideration in this study. Uncertainties in the prediction will be reduced when the natural convection and ablation phenomena are more rigorously dealt with in the code, and if more accurate initial and time-dependent conditions are supplied from the test in terms of material composition and its associated thermophysical properties.

• Nuclear Engineering and Technology
• /
• v.52 no.1
• /
• pp.185-191
• /
• 2020

Plant-specific analyses of an advanced reactor have been performed to assure the structural integrity of the reactor pressure vessel during transient conditions, which are expected to initiate pressurized thermal shock (PTS) events. The vessel failure probabilities from the probabilistic fracture mechanics analyses are combined with the transient frequencies to generate the through-wall cracking frequencies, which are compared to the acceptance criterion. Several sensitivity analyses are performed, focusing on the orientations and sizes of cracks, the copper content, and a flaw distribution model. The results show that the integrity of the reactor vessel is expected to be maintained for long-term operation beyond the design lifetime from the PTS perspective using the design data of the advanced reactor. Moreover, a fluence level exceeding 9×10¹⁹ n/㎠ is found to be acceptable, generating a sufficient margin beyond the design lifetime.

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

• Nuclear Engineering and Technology
• /
• v.52 no.2
• /
• pp.448-455
• /
• 2020

The H.B. Robinson Unit 2 (HBR-2) pressure vessel dosimetry benchmark is an in- and ex-Reactor Pressure Vessel (RPV) neutron dosimetry benchmark based on experimental data from the HBR-2 reactor, a 2300-MW PWR designed by Westinghouse and put in operation in March 1971, openly available through the SINBAD Database at OECD/NEA data Bank. The goals of the present work were to carry out three-dimensional (3D) fixed source transport calculations in both Cartesian (X,Y,Z) and cylindrical (R,θ,Z) geometries by using the TORT-3.2 discrete ordinates code on very detailed 3D HBR-2 geometrical models and to test the latest broad-group coupled (47 neutron groups + 20 photon groups) working cross section libraries in FIDO-ANISN format with same structure as BUGLE-96, such as BUGJEFF311.BOLIB, BUGENDF70.BOLIB and BUGLE-B7. The results obtained with all the cited libraries were satisfactory and are here reported and compared.

• Proceedings of the KSME Conference
• /
• 2001.06a
• /
• pp.441-446
• /
• 2001

In the integral type reactor, SMART, all the major components such as steam generators, pressurizer and pumps are located inside the single reactor pressure vessel. The objective of this study is to evaluate the structural integrity for RPV of SMART under the postulated pressurized thermal shock by applying the finite element analysis. Input data for the finite element analysis were generated using the commercial code I-DEAS, and the fracture mechanics analysis was performed using the ABAQUS. The crack configurations, the crack aspect ratio and the clad thickness were considered in the parametric study. The effects of these parameters on the reference nil-ductility transition temperature were also investigated.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.24 no.11
• /
• pp.2866-2874
• /
• 2000

This paper is concerned with the optimum design for external reinforcement 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 severs accident. The reis of the temperature and the internal pressure cause deterioration of the load carrying capacity and could cause failure of the RPV lower head. The deterioration of failure can be mitigated by the external cooling or the reinforcement of the lower head with additional structures. While the external cooling forces the temperature of an RPV to drop to the desired level, the reinforcement of the lower head can attain both the increase of the load carrying capacity and the temperature drop. The reinforcement of the lower head can be optimized to have the maximum effect on the collapse pressure and the temperature at the inner wall. Optimization results are compared to both the result without the reinforcement and the result with the designated reinforcement.

• Nuclear Engineering and Technology
• /
• v.53 no.9
• /
• pp.3112-3121
• /
• 2021

One of the major factors affecting the life span of a Reactor Pressure Vessel (RPV) is the Pressurised Thermal Shock (PTS). PTS is a thermo-mechanical load on the RPV wall due to steep temperature gradients and structural load created by internal pressure of the fluid within the RPV. Safe operating life of a nuclear power plant is ensured by carrying out fracture analysis of the RPV against thermal shock. Carrying out fracture tests on RPV/large scale components is not always feasible. Hence, studies on laboratory level specimens are necessary to validate and supplement the prototype results. This paper aims to study the fracture behaviour of standard Compact Tension [C(T)] specimens, made of RPV steel 20MnMoNi55, subjected to thermal shock through experimental and numerical investigations. Fracture tests have been carried out on the C(T) specimens subjected to thermal transient load and tensile load to quantify the effect of thermal shock. Crack resistance curves are obtained from the fracture tests as per ASTM E1820 and compared with those obtained numerically using XFEM and a good agreement was found. A quantitative study on the crack tip plastic zone, computed using cohesive segment approach, from the numerical analyses justified the experimental crack initiation toughness.

• 한국전산유체공학회:학술대회논문집
• /
• 2009.11a
• /
• pp.121-128
• /
• 2009

During a hypothetical high-pressure accident in a nuclear power plant (NPP), molten corium can be ejected through a breach of a reactor pressure vessel (RPV) and dispersed by a following jet of a high-pressure steam in the RPV. The dispersed corium is fragmented into smaller droplets in a reactor cavity of the NPP by the steam jet and released into other compartments of the NPP by a overpressure in the cavity. The fragments of the corium transfer thermal energy to the ambient air in the containment or interact chemically with steam and generate hydrogen which may be burnt in the containment. The thermal loads from the ejected molten corium on the containment which is called direct containment heating (DCH) can threaten the integrity of the containment. DCH in a NPP containment is related to many physical phenomena such as multi-phase hydrodynamics, thermodynamics and chemical process. In the evaluation of the DCH load, the melt dispersion rates depending on the RPV pressure are the most important parameter. Mostly, DCH was evaluated by using lumped-analysis codes with some correlations obtained from experiments for the dispersion rates. In this study, MC3D code was used to evaluate the dispersion rates in the APR1400 NPP during the high-pressure accidents. MC3D is a two-phase analysis code based on Eulerian four-fields for melt jet, melt droplets, gas and water. The dispersion rates of the corium melt depending on the RPV pressure were obtained from the MC3D analyses and the values specific to the APR1400 cavity geometry were compared to a currently available correlation.

• Corrosion Science and Technology
• /
• v.18 no.5
• /
• pp.190-195
• /
• 2019

SA508 low-alloy steel for a reactor vessel was exposed to primary water in a pressurized water reactor (PWR) plant because the cladding layer of type 309 stainless steel for the RPV was removed, due to an accident in which the detachment of the thermal sleeve occurred. The major advantage of the electrochemical deposition (ECD) Ni plating technique is that the reactor pressure vessel can be repaired without significant thermal effects, and Ni has solid corrosion resistance that can withstand boric acid. The corrosion rate assessment of the damaged part was performed, and its trend was analyzed. Essential variables of the Ni plating for repair of the damaged part were derived. These conditions are applicable variables for the repair plating device, and have been carefully adjusted using the repair plating device. The process for establishing ASME technical standards called Code Case N-840 is described. The process of developing Ni-plating devices, and the electroplating procedure specification (EPS) are described.

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1651-1664
• /
• 2023

It is very difficult to capture the multi-dimensional phenomena such as asymmetric flow and temperature distributions with the one-dimensional (1D) model, obviously, due to its inherent limitation. In order to overcome such a limitation of the 1D representation, many state-of-the-art system codes have equipped a three-dimensional (3D) component for multi-dimensional analysis capability. In this study, a standard multi-dimensional analysis model of APR1400 (Advanced Power Reactor 1400) has been developed using TRACE (TRAC/RELAP Advanced Computational Engine). The entire reactor pressure vessel (RPV) of APR1400 has been modeled using a single 3D component. The fuels in the reactor core have been described with detailed and coarse representations, respectively, to figure out the impact of the fuel description. Using both 3D RPV models, a comparative analysis has been performed postulating a double-ended guillotine break at a cold leg. Based on the results of comparative analysis, it is revealed that both models show no significant difference in general plant behavior and the model with coarse fuel model could be used for faster transient analysis without reactor kinetics coupling. The analysis indicates that the asymmetric temperature and flow distributions are captured during the transient, and such nonuniform distributions contribute to asymmetric quenching behaviors during blowdown and reflood phases. Such asymmetries are directly connected to the figure of merits in the LBLOCA analysis. Therefore, it is recommended to employ a multi-dimensional RPV model with a detailed fuel description for a realistic safety analysis with the consideration of the spatial configuration of the reactor core.