G. Bartholomé

Assessment of large scale pipe tests by fracture mechanics approximation procedures with regard to leak-before-break

Abstract On the basis of numerous tests which were performed to demonstrate the “Leak-Before-Break” (LBB) behaviour predictions based on analyses could be further developed and substantiated. The methods applied (local flow stress concept, plastic limit load concept and two criteria approach) for the calculation of failure loads in pipes with circumferential cracks subjected to bending moment and internal pressure loading are evaluated and compared to the results gained by experiments. LBB predictions on the basis of the methods described show that reliable statements can be made. This has been confirmed by the experiments performed.

Preclusion of double-ended circumferential rupture of the main coolant line

Abstract The basic safety concept ensures a quality standard of the reactor coolant pressure boundary which precludes a catastrophic failure caused by defects in the manufacture. Circumferential ruptures need no longer be postulated for the main coolant line due to the reduction of the stresses occurring, e.g. by optimization of the mechanical design, the markedly tough condition of the materials and the high quality of manufacture and processing.

Linear elastic stress intensity factors for cracks in nuclear pressure vessel nozzles under pressure and temperature loading

Abstract Stress intensity factors for two different nozzle geometries and different crack sizes are evaluated for pressure and thermal loading utilising three-dimensional elastic finite element models. The results are compared to available experimental data and a procedure is proposed to estimate the maximum of the stress intensity factor for arbitrary crack size and loading conditions.

Design and calibration of leak detection systems by thermal hydraulics and fracture mechanics analyses

Abstract Leak-before-break behaviour can be proven if a leak, which occurs after a crack has penetrated the wall, can be safely detected by a leak detection system long time before a critical crack size is reached. For the efficiency of leakage detection systems the knowledge of the correlation between leakage detection signals (temperature, condensate, humidity, acoustic parameters) and leakage rates, leakage areas and through-wall crack lengths is required. The leakage area is computed with a fracture mechanics model using the integration of the crack opening along the crack front, considering plasticity and geometrical effects. The calculation of the leak rate from a through-wall crack in a component is based on a thermal hydraulics model taking into account the effects of pressure, temperature, crack length, crack width, leakage area, crack depth (wall thickness), roughness and hydraulic coefficient of resistance. These analytical approaches, both fracture mechanics and thermal hydraulics, are used to quantify the availability and safety margins of components in connection with plant life extension analyses. The application of the above procedure is shown for the main coolant line of a SIEMENS-PWR (MCL) and the surge line of a WWER-440-PWR (SL).

Determination of critical circumferential through-wall crack sizes regarding load reduction by increased flexibility in piping systems

Piping systems in power stations are not only loaded by the system pressure and force controlled bending moments e.g. due to the weight of the system, but also by additional bending moments caused by the thermal expansion of the hot system and prescribed motions, as in the case of an earthquake, resulting in displacement controlled moments. To determine critical circumferential through-wall crack sizes in pipes, a method based on the J-integral has been developed, which takes into account the load reduction due to the increased flexibility of a flawed section. The method is presented and compared with results by the finite element method and experimental data.

Activities in the field of plant life evaluation, life extension and plant improvement

Abstract The adequate margin of safety reached in the as-built condition for a nuclear power plant shall be maintained throughout the whole life of the plant. To attain this, a systematic lifetime evaluation of safety related items should be performed in due time in the light of new developments. The results can be used for the purpose of life extension and license renewal, too. Siemens has gained an integral analysis concept by practical experience. The analysis can result in corrective actions with respect to life extension of plant components and systems or can lead to measures for plant improvement. Examples of performed activities are given within this article.

Validation of simplified fracture mechanics methods by testing of real components

Abstract The evaluation of integrity of structural components is often based on the proof of leak-before-break (LBB). Leak-before-break behaviour in piping constitutes a fail-safe condition. Which means that, during multiplied loading conditions, a defect results at first in a leakage. The crack length which leads to the leakage is smaller than the critical through-wall crack length. Simplified fracture mechanics concepts are used for the demonstration of LBB. For this the conservative, safe calculation of the critical through-wall crack length for ductile failure is necessary. To validate simplified calculation methods for circumferential cracks (flow stress concept (FSC); plastic limit load (PLL)) and for axial cracks (Battelle approach (BMI); Ruiz approach (RUIZ)) all available experiments on real structural components, especially on pipes, were analysed and evaluated by the mentioned simplified methods (approximately 460 experiments). The methods were adapted by application of correction factors, mainly on the flow stress, to result in conservative (safe) and realistic (as near as possible to the experiments) predictions. Depending on method (FSC, PLL, BMI, RUIZ), crack orientation (circumferential and axial cracks) and type of material (ferritic and austenitic material) different definitions of flow stresses were established.

Critical circumferential through-wall cracks according to the unloading of the cracked section under displacement-controlled bending load

Abstract The evaluation of critical circumferential through-wall crack lengths in piping is usually performed by the flow stress concept, plastic limit load method or GE-EPRI procedures. Most of these methods treat the secondary stresses, especially those caused by bending moments resulting from restrained thermal expansion, as force-controlled loads. In reality, there is a movement of the piping into the direction of the prescribed displacement and, therefore, a relaxation of the cracked section, which is due to the local rotation of the cracked section. Instead of the bending moment originating from the elastic analysis of the piping system there will be a reduced bending moment, the load decreases, the real critical through-wall crack lengths due to this displacement-controlled loading are larger than those predicted by the load controlled methods. A corresponding analytical procedure taking into account this relaxation was developed and validated by a comparison with experiments as well as finite element calculations. The procedure can be used for the evaluation of the safety of piping systems (e.g. leak-before-break analyses), if the usual methods based on force-controlled loads give unrealistic, conservative results.

Safety analysis of reactor pressure vessels considering irradiation effects and improved temperature calculations for PTS

Abstract The safety analysis of reactor pressure vessels has to take into account all parameters, including design, material, fabrication, inspection, loading and service conditions. This safety analysis for pressurized thermal shock consists of the following: materials analysis (toughness, irradiation); defect analysis (inspection during fabrication and while in service); thermal hydraulic analysis (fluid-fluid mixing scenario, fluid-fluid mixing experiments); fracture mechanics analysis (evaluation of material, defects and loads). The practice with respect to this procedure is described in this paper.

Sicherheitsanalyse von Reaktordruckbehältern

Abstract An evaluation of the safety of nuclear reactor pressure vessels for pressurized water reactors to resist fracture is an important feature in ensuring nuclear reactor safety. This evaluation is performed using fracture mechanics. Basic theory of fracture mechanics The linear, elastic and plastic stress distributions for plane stress and plane strain conditions are derived for the different modes of fracture. The concepts of the linear elastic fracture mechanics and of the crack-opening method are shown. A critical analysis of crack growth and of stress corrosion cracking extends the use of fracture mechanics. Technical application of fracture mechanics on nuclear reactor pressure vessel: Design and calculation of reactor pressure vessels using brittle fracture concepts; estimation of permissible crack sizes for the specification of standards for inspection, hydrostatic pressure test and recurring inspections; evaluation of the operating performance and life expectancy of nuclear reactor pressure vessels including crack growth and stress corrosion cracking. Safety evaluation Using the above derived data, the safety of reactor pressure vessels against fracture is shown by means of: crack size-stress diagrams; brittle fracture diagrams and ductile fracture diagrams; operating stress diagrams.