Unstable crack propagation under severe accident scenario conditions in a pressurized water reactor (original) (raw)
Related papers
Nuclear Engineering and Design, 1995
A joint pressure vessel integrity research programme involving three partners is being carried out during 1990-1995. The partners are the Central Research Institute of Structural Materials "Prometey" from Russia, IVO International Ltd (IVO) from Finland, and the Technical Research Centre of Finland (VTT). The main objective of the research programme is to increase the reliability of the VVER-440 reactor pressure vessel safety analysis. This is achieved by providing material property data for the VVER-440 pressure vessel steel, and by producing experimental understanding of the crack behaviour in pressurized thermal shock loading for the validation of different fracture assessment methods. The programme is divided into four parts: pressure vessel tests, material characterization, computational fracture analyses, and evaluation of the analysis methods. The testing programme comprises tests on two model pressure vessels with artificial axial outer surface flaws. The first model vessel had circumferential weld seam at the mid-length of the vessel. A special embrittling heat treatment is applied to the vessels before tests to simulate the fracture toughness at the end-of-life condition of a real reactor pressure vessel. The sixth test on the first model led to crack initiation followed by arrest. After the testing phase, material characterization was performed. Comparison of calculated and experimental data generally led to a good correlation, although the work is being continued to resolve the discrepancies between the measured initiation and arrest properties of the material. 0029-5493/95/$09.50 © 1995 Elsevier Science S.A. All rights reserved SSDI 0029-5493(95)01030-0
Deterministic assessment of reactor pressure vessel integrity under pressurised thermal shock
International Journal of Pressure Vessels and Piping, 1998
Numerical investigations were carried out to assess the integrity of reactor pressure vessels under pressurised thermal shock (PTS). The 4loop reactor pressure vessel with cladding was subjected to thermo-mechanical loading owing to loss of coolant accident. The loss of coolant accident corresponding to small break as well as hot leg breaks were considered separately, which led to axisymmetric and asymmetric thermal loading conditions respectively. Three different crack configurations, 360Њ circumferential part through, circumferential semielliptical surface and circumferential semi-elliptical under-clad cracks, were postulated in the reactor pressure vessel. Finite element method was used as a tool for transient thermo-elastic analysis. The various fracture parameters such as crack mouth opening displacement (CMOD), stress intensity factor (SIF), nil ductility transition temperature (RT NDT ) etc. were computed for each crack configuration subjected to various type of loading conditions. Finally for each crack a fracture assessment was performed concerning crack initiation based on the fracture toughness curve. The required material RT NDT was evaluated to avoid crack initiation. ᭧
Fracture mechanical evaluation of an in-vessel melt retention scenario
Annals of Nuclear Energy, 2008
This paper presents methods to compute J-integral values for cracks in two-and three-dimensional thermo-mechanical loaded structures using the finite element code ANSYS. The developed methods are used to evaluate the behavior of a crack on the outside of an emergency cooled reactor pressure vessel (RPV) during a severe core melt down accident. It will be shown, that water cooling of the outer surface of a RPV during a core melt down accident can prevent vessel failure due to creep and ductile rupture. Further on, we present J-integral values for an assumed crack at the outside of the lower plenum of the RPV, at its most stressed location for an emergency cooling (thermal shock) scenario.
Nuclear Technology, 2020
In recent years, the compound beyond-design-basis accident (BDBA), which combines earthquake, tsunami, or some other severe events to impact a nuclear power plant (NPP), has received more attention. After the Fukushima nuclear disaster, the licensee of NPPs in Taiwan established the ultimate response guideline (URG) that instructs operators to perform reactor depressurization, low-pressure water injection, and containment venting to prevent core meltdown and hydrogen explosion once long-term loss-of-power and water-supply events occur. In this paper, we employed the probabilistic fracture mechanics (PFM) method to evaluate the structural integrity of boiling water reactor (BWR) pressure vessels under URG operation. At first, models of the beltline shell welds for BWR vessels associated with the Pressure Vessel Research Users Facility-Exponential flaw distribution were built for the PFM Fracture Analysis of Vessels-Oak Ridge (FAVOR) code. Then, the thermal-hydraulic data of URG transients for Taiwan domestic BWRs were imposed as the loading conditions. The analysis results demonstrate that performing URG operation will not cause significant fracture probability even at extreme embrittlement conditions. If longterm station blackout occurs due to a compound BDBA, the URG procedures can prevent core damage and hydrogen explosion, while maintaining the structural integrity of the reactor pressure vessels.
Probabilistic assessment of reactor pressure vessel integrity under pressurised thermal shock
International Journal of Pressure Vessels and Piping, 1999
A deterministic fracture mechanics analysis does not address the uncertainties involved in material properties, magnitudes of loads, location and size of the flaws, etc. However, in a real life situations such uncertainties can affect significantly the conclusions drawn out of a deterministic analysis. The principles of probabilistic fracture mechanics may be used to ascertain the effects of such uncertainties. A computer code PARISH (Probabilistic Assessment of Reactor Integrity under pressurised thermal SHock) has been developed based on principles of PFM for analysing a reactor vessel subjected to pressurised thermal shock. The code assumes a crack in the reactor vessel of random dimension depending upon Marshall flaw depth cumulative distribution function. The applied SIF at the tip of this crack is computed either using closed form solution or a precomputed data base. The material K IC is then calculated using the crack tip temperature and RT NDT. The value of RT NDT depends on the initial value of RT NDT and the increase in the value of RT NDT depending upon the fluence, copper content and nickel content. A Gaussian distribution is assumed for these parameters. If the applied SIF is more than the material K IC , the crack is assumed to propagate. The crack can be arrested only if the applied SIF is less than the material K Ia at that location. The material K Ia again depends upon the RT NDT, which in turn depends upon the fluence, copper content and nickel content of the material at that location. The vessel failure is assumed if the crack propagates by the 75% of the thickness. Such procedure is repeated for large number of cracks (of the order of one million). Using Monte-Carlo simulation, probabilities of no crack initiation, crack initiation and vessel failure are calculated. The present probabilities are conditional in the sense that the transient is assumed to occur. The case studies are presented involving a nuclear reactor vessel subjected to two different kinds of pressurised thermal shocks.
Estimation of crack opening area for leak before break analysis of nuclear reactor system
Nuclear Engineering and Design, 2009
The crack opening area has been estimated by various models for leak before break analysis of nuclear reactor systems. A number of linear elastic fracture mechanics models have been employed. A model has been derived based on Forman et al. [Forman, R.G., Hickman, J.C., Shivakumar, V., 1985. Stress intensity factors for circumferential through cracks in hollow cylinders subjected to combined tension and bending loads. Eng. Frac. Mech. 21, 563-571] model for stress intensity factors. The linear elastic fracture models entail the use of plasticity corrections. The convergence behaviour of the iterative implementation for plasticity correction has been studied. An elasto-plastic fracture mechanics model has also been adopted for comparative study undertaken here. The range of applicability of the models has been reviewed.
Leak-Before-Break Assessment of a Cracked Reactor Vessel Nozzle Using 3D Crack Meshes
Volume 6A: Materials and Fabrication, 2016
A postulated surface crack near a reactor pressure vessel nozzle is evaluated using finite element analysis (FEA) to compute the fatigue crack growth rate, evaluate crack stability, and examine the possibility of a leak-before-break (LBB) condition. For a pressurized vessel with cyclic loading, determining if the crack may have a LBB condition is desirable to allow for the possibility of leak detection leading to corrective action before catastrophic failure. A fatigue crack growth analysis is used to determine how the surface crack dimensions develop before re-categorizing the surface crack as a through thickness crack and evaluating its stability for LBB. To evaluate if a particular crack is unstable and may cause a structural failure, the Failure Assessment Diagram (FAD) method provides an evaluation using two ratios: brittle fracture and plastic collapse. The FAD method is described in the engineering best practice standard API 579-1/ASME FFS-1. The FAD curve and assessment ratios can be obtained from crack front J-integral values, which are computed using 3D crack meshes and elastic and elastic-plastic FEA. Computing custom crack solutions is beneficial when structural component geometries do not have an available stress intensity or reference stress solution.
Structural integrity assessment of reactor pressure vessels during pressurized thermal shock
Journal of Mechanical Science and Technology, 2008
A comparative assessment study is performed for the deterministic fracture mechanics approach of the pressurized thermal shock of a reactor pressure vessel. Round robin problems consisting of two transients and two defects are solved. Their results are compared to suggest some recommendations of best practices and to assure an understanding of the key parameters of this type of approach, which will be helpful not only for the benchmark calculations and results comparisons but also as a part of the knowledge management for the future generation. Seven participants from five organizations solved the problem and their results are compiled in this study.