Integrity Evaluation of a Reactor Pressure Vessel Based on a Sequential Abaqus-FRANC3D Simulation Method (original) (raw)
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Science and Technology of Nuclear Installations, 2022
e damage induced pressurized thermal shock (PTS) may pose to a reactor pressure vessel (RPV) is a critical safety requirement assessed as part of the ageing management programme of pressurized water reactors (PWRs). A number of researches have studied PTS initiated mainly by postulated accidents such as loss of coolant accidents (LOCAs). However, investigations on PTSinduced threat on RPV caused by inadvertent actuation of the safety injection, a frequent anticipated transient, have not been thoroughly studied. In this paper, a simplified multistep analysis method is applied to study the thermomechanical status of a twoloop PWR under PTS loads caused by inadvertent actuation of the safety injection system. A direct-coupling thermomechanical analysis is performed using a three-dimensional (3D) RPV finite element model. A 3D finite element submodel (consisting of the highiest stress concentration area in the RPV) and an assumed crack are then used to perform fracture mechanics analysis. Subsequently, the critical integrity parameter-stress intensity factor (SIF) is estimated based on FRANC3D-M-integral method coupled in the multistep simulation. e material fracture toughness of the vessel is computed based on the master curve method with experimental fracture toughness data. e results obtained from the direct coupling stress analysis in comparison with sequential coupling approach demonstrate the effectiveness of the proposed multistep method. Also, comparing SIF results obtained with that calculated based on the conventional virtual crack-closure technique (VCCT) and extended finite element method (XFEM) show good agreement. is study provides a useful basis for future studies on anticipated transient-induced crack propagation and remaining service life prediction of ageing reactor pressure vessels.
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. ᭧
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.
Determining Useful Life of Reactor Pressure Vessels
The length of useful life of the reactor vessel is much more questionable than the estimate of a probable breakdown. It particularly arises as a need and an obligation after regular technical inspections. The problem of determining the remaining life of the vessel is especially arises when there are errors or cracks on the vessel walls. Research on high quality vessels leads to the conclusion that hydraulic tests are harmful, which was confirmed by a large number of cracks that appeared after these hydraulic tests. This fact gives a basis for evaluation of these tests and their adjustment depending on the length of exploitation. The analysis of causes for a change in material properties from manufacturing all the way to the removal from exploitation was performed on over fifty vessels in oil processing. Special attention was given to a reactor vessel that operates under heavy operating conditions and with a complex catalytic chemical process using hydrogen. Changes in material characteristics were monitored from the production of the reactor vessel until its shutdown and removal from service. The causes for crack occurrence and expansion in vessel walls and heads were analyzed. The occurrence of new cracks and expansion of old cracks was particularly considered during hydraulic tests.
Analysis of a reactor pressure vessel subjected to pressurized thermal shocks
International Journal of Computational Methods and Experimental Measurements
The integrity of reactor pressure vessels (RPVs) of nuclear power plants is one of the most important topics in the field of nuclear energy production. Therefore, the integrity of RPVs has to be assessed for normal operation as well as for emergency transients. A critical transient concerning the RPV integrity is the emergency cooling of a pressurized water reactor, initiated by a leak in the hot leg. Such shock-like cooling in combination with the pressure, the so-called pressurized thermal shock (PTS), causes high thermal stresses in the RPV wall and stress intensities of pre-existing cracks which could exceed the remaining fracture toughness of the material, which is additionally embrittled due to neutron irradiation. This may result in a cleavage fracture of the most safety relevant reactor component. We present a PTS study of a reference reactor, starting with the calculation of the thermal-hydraulic system behaviour, followed by the simulation of the cold water temperature injection and mixing by means of computational fluid dynamics (CFD) method and the subsequent structural and fracture mechanics calculation. In the safety assessment, we compare the evolution of the stress intensity factors (SIF) during an emergency cooling transient with the fracture toughness at the tip of postulated cracks. Results and open questions will be discussed in the light of a realistic estimation of safety margins.
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.
Engineering Fracture Mechanics, 2013
The integrity analysis of a reactor pressure vessel subjected to pressurized thermal shocks is performed. Linear elastic analysis leads to a more conservative result than the elasticplastic analysis if the warm prestressing effect is not considered. The stress intensity factor for the deepest point of a surface crack front is not always larger than that for a surface point, indicating that both the deepest and surface points of the crack front should be considered. The safety margin of the reactor pressure vessel is larger based on the K-T approach than that only based on a K approach.
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.
International Journal of Pressure Vessels and Piping, 2008
The analysis of the stability of a defect in a cladded reactor pressure vessel (RPV) of a nuclear pressure water reactor (PWR) subjected to pressurised thermal shock (PTS) is one main elements of the general safety demonstration. Recently, CEA proposed several improved analytical tools for the analysis of the PTS. First, an analytical solution for the vessel through-thickness temperature variation has been developed to deal with any fluid temperature, taking into account the possible presence of a cladding, in the case of an internal PTS. The associated thermal stress expression has been simplified and a complete linearised solution is given for the thermal loading and also for internal pressure, depending on the main vessel material and on the cladding properties. Finally, a complete compendium is also given for the elastic stresses intensity factor calculation. This paper proposes several improvements of the proposed analytical method to deal with a PTS in a PWR cladded vessel. A variable heat transfer coefficient is now taken into account based on an equivalent fluid temperature variation determination, associated with a constant heat transfer coefficient, to keep the same thermal exchange between the fluid and the inner skin of the vessel obtained with the initial data. A more accurate expression for the linearised stresses due to the internal pressure is given, and a possible effect of residual stresses due to the difference between the operating temperature and the stress-free temperature is also taken into account. Finally, an extension of the domain of definition of the influence functions for the elastic stress intensity factor calculation is given.