Kari Törrönen

On the mechanisms of environment sensitive cyclic crack growth of nuclear reactor pressure vessel steels

Abstract An analysis of the cyclic crack growth rate data generated so far for pressure vessel materials in simulated light water reactor environments suggests a strong dependence on frequency, load ratio, waveform, temperature and material composition. To account for these observations a mechanistic crack growth model has been advanced based on hydrogen-induced cracking, where anodic dissolution creates the conditions for hydrogen absorption at the crack tip. Hydrogen-induced cracking starts from the manganese sulfide inclusions, which act as strong hydrogen traps. The hydrogen-induced crack growth around the manganese sulfide inclusions generally spans several prior austenite grains. At high hydrogen input rates brittle crack growth also occurs, this being unrelated to inclusions. When the crack growth exposes manganese sulfide inclusions, these dissolve and the crack tip environment becomes aggressive and conducive to hydrogen absorption. This hydrogen-induced cracking model explains why inclusions form a preferred crack path, and accounts for the effect of sulfur on crack growth rate both in PWR and BWR conditions. Based on the model, the observed crack growth rate dependence on different testing variables can also be explained.

Environment sensitive cracking in pressure boundary materials of light water reactors

Abstract A review of the various forms of environment sensitive cracking in pressure boundary materials of light water reactors is presented. The available methods and the most promising future possibilities of preventive maintenance to counteract the environmental degradation are evaluated. Environment sensitive cracking is considered from the metallurgical, mechanical and environmental points of view. The main emphasis is on intergranular stress corrosion cracking of austenitic stainless steels and high strength Ni-base alloys as well as on corrosion fatigue of low alloy and stainless steels. Additionally, some general ideas on how to predict, reduce, monitor or eliminate environment sensitive cracking in service are presented.

Theory based statistical interpretation of brittle fracture toughness of reactor pressure vessel steel 15X2MφA and its welds

Abstract The reliability of safety analysis of nuclear reactor pressure vessels is imperatively dependent upon the reliability of the fracture toughness values used in the analysis. Normally, existing fracture toughness results are analyzed with standard statistical means making use of empirically defined distribution functions for the fracture toughness. Recent developments in the theoretical modelling of cleavage fracture initiation have evolved a sounder statistical description of the macroscopic fracture toughness. It has thus become possible to make a theoretical interpretation of the scatter in brittle fracture toughness results. Here, the theoretical relations are applied to the statistical analysis of brittle fracture toughness results for the reactor pressure vessel steel 15X2MφA and its welds. Based on the analysis, theoretical statistical reference curves for the steels are presented and compared to the presently used standard reference curves.

Optimization of metallurgical variables in fracture prevention

Abstract The effect of metallurgical variables on the two most important crack growth mechanisms - stable crack growth by environmentally assisted cyclic crack growth and unstable crack growth by cleavage - in light water reactor pressure vessel steels is evaluated. The analyses are based on micromechanisms of fracture and sensitivity analysis, when applicable. Metallurgical variables considered are non-metallic inclusions and carbides as well as other parameters through their effects on yield strength and other mechanical properties.

Fracture behaviour of large scale pressure vessels in the hydrotest

Abstract Within the Nordic Countries a four-year research programme in the area of elastic-plastic fracture mechanics was initiated in 1985. This programme aims to assess the leak-before-break (LBB) criteria for pressure vessels and piping. The main experimental effort of the programme is to rupture large size pressure vessels, one having dimensions resembling those of a nuclear reactor pressure vessel, under internal pressure. Artificial flaws were made on the inner wall of the vessels. The dimensions of the flaws were defined by calculations so that the LBB condition was just anticipated during the test. For the time being two tests have been performed. The first test with a large pressure vessel was pressurized by water at 60°C, which was the lowest acceptable temperature for the hydrotest. In this paper experimental details including flaw preparation, instrumentation and material characterization are described. The fracture behaviour as well as experimental results of the tests are reported and compared to the analytical solutions of the analyses.

Evaluation of the effect of metallurgical variables on materials behaviour and reference curves

Abstract An integral part of the safety assessment of nuclear pressure vessels and piping is the quantitative estimation of defect growth in both a stable and an unstable manner during service. This estimation is essential for determining whether any defect detected during inspection should be repaired or whether the size of the defect even after its expected growth is small enough to leave the integrity of the vessel unaffected. The most important stable defect growth mechanism is that of environmentally assisted cyclic crack growth. Recent results indicate that it is markedly affected by sulphur content and/or manganese sulphide morphology and distribution. This implies that an essential improvement in component safety has been gained by currently applied steelmaking practices, which result in extra low sulphur content, generally below 0·010 wt.%, and in the round shape and small size of inclusions through, e.g. calcium treatment, hence considerably reducing the effect of the environment on crack growth rate. This further implies that the ASME Section XI reference curves for environmentally accelerated cyclic crack growth are conservative for steels produced by current steelmaking practices. The ASME Section XI applies predominantly linear elastic fracture mechanics to assess the effects of cracks on the integrity of nuclear power plant components. Unstable linear elastic fracture often propagates by a cleavage mechanism. The cleavage fracture process has recently been shown to be of a statistical nature in both ferritic and bainitic steels. The carbide size distribution plays a dominant role in controlling the fracture toughness of these steels. A cleavage fracture model has been developed, by which both the expectance value and the probability limits of the fracture toughness, K IC , can be predicted. The probability limits given by the model are shown to be consistent with the experimental observations. The application of the model to the data on which the ASME Section XI reference fracture toughness curve is based indicates that the reference curve is slightly unconservative.

Crack initiation and arrest in a pressurized thermal shock test for a model pressure vessel made of VVER-440 reactor pressure vessel steel

Abstract 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.

Comparison of Charpy-V and J-integral transition temperature shifts in CrMoV pressure vessel steels

Abstract Both Charpy-V and precracked Charpy-size specimens were included in the original surveillance programme of the Loviisa nuclear power plant pressure vessel made of CrMoV steel. With precracked specimens the J- value at the onset of brittle fracture was determined as a function of temperature and irradiation dose. The irradiation induced transition temperature shifts measured with Charpy-V and precracked three-point bend specimens are correlated by using indexed temperature values. The shapes of brittle fracture transition curves measured with the fracture mechanical specimens both for the irradiated and non-irradiated material state are compared using a statistical brittle fracture model.

Fracture behaviour simulation of flawed full scale pressure vessel

Abstract In 1985 a four year research programme in elastic-plastic fracture mechanics was initiated within the Nordic countries. The aim of the programme was to verify the methods used for fracture analysis of real structures. A large cylindrical pressure vessel, having dimensions resembling those of a nuclear reactor pressure vessel, was tested in 1986. An artificial sharp axial surface flaw was made on the inner wall of the vessel. One of the circumferential welds intersected the crack at its midpoint. Failure of the vessel occurred as local rupture in the weld area. A maintenance deck, which was located around the midsection of the pressure vessel, was partially removed so as to prevent interference with the expected vessel deformations upon pressurization. After the test, two three-dimensional nonlinear finite element analyses were performed taking into account the existence of the circumferential weld in the ligament. In the first case, the original flawed structure was modelled allowing for different stress-strain curves of the base and weld material. In the second analysis, the model included a short through-the-wall crack in the weldment. Additionally, a simple two-dimensional analysis was made assuming the crack to be infinitely long in the axial direction. A three-dimensional analysis was repeated without considering the effect of the maintenance deck. For fracture mechanics evaluation, J- integrals along the crack front were calculated. In this paper, results of the three-dimensional analyses are reported and compared to experimental findings.

On thermal annealing of irradiated PWR pressure vessels

Radiation embrittlement in some pressurised water reactors has been so fast that, in spite of other applied mitigation methods, thermal annealing has been practically the only solution permitting further operation. The annealings have been reported to be successful and resulted in no damage. In cases where the whole fuel core zone area of the reactor pressure vessel has to be annealed, a fully successful annealing has yet to be convincingly proven. High thermal stresses may make the thermal treatment troublesome to carry out. The recovery mechanisms have been difficult to study because of the exceptionally small size of irradiation defects. The degree of recovery cannot be yet fully calculated precisely from material and annealing information. Especially the recovery processes in restoring elastic and elastic-plastic fracture toughness properties needs much more work to be carried out in order to be clarified. Recovery annealing at a proper temperature is, however, a very effective method and, in many cases, practically the only alternative for extending the service life of a pressure vessel embrittled by radiation. There seems to be no restrictions to repeating the thermal treatment, but every pressure vessel should be independently studied and assessed for the achievement of safe results.