Fracture mechanics evaluation of the structural material for fast breeder reactors

Results of fracture mechanics investigations on austenitic steels used for LMFBRs (Liquid Metal Fast Breeder Reactors) are presented. A summary of reported tests on straight piping and elbows with through wall flaws is given which agree well with predictions made by using a plastic instability model. Crack growth experiments and calculations indicate that initial flaws will not extend significantly during service. Even if considerable crack growth is postulated cracks will penetrate the piping wall with a high safety margin to unstable crack configurations. Theoretical investigations of flawed structures under high strains show that the effect of crack size can be discussed similarly to the elastic range. The information demonstrate that with respect to the design requirements and operating conditions of LMFBRs a sudden rupture of the piping can be excluded. The integrity of the coolant boundary is given also in case of initial flaws.     

Fracture mechanics concepts for evaluation of irradiated reactor pressure vessel material behavior

Within the German research program Forschungsvorhaben Komponentensicherheit (FKS), irradiation experiments were performed with ferritic reactor pressure vessel (RPV) steels and welds. The materials cover a wide range of chemical composition and initial toughness to achieve different susceptibility to neutron irradiation. Different neutron flux was applied and the neutron exposure extended up to 8×10¹⁹ cm⁻². The change in material properties was determined by means of tensile, Charpy impact, drop-weight and fracture mechanics tests, including crack arrest. The results have provided more insight into the acting embrittlement mechanisms and shown that the fracture mechanics concept of the Code provides in general an upper bound for the material which can be applied in the safety analysis of the RPV.     

Fracture mechanics issues for irradiated blanket structures

This paper reviews some of the factors that will affect fracture behavior of fusion reactor structures and summarizes some component life predictions based on linear elastic fracture mechanics analysis. The review includes discussion of the environments to which the components will be subjected, the response of materials to these environments, the time dependent nature of the structural response, and the fracture related failure mechanisms.Radiation environments and complex loading conditions in a fusion reactor cause a variety of material phenomena. These phenomena include irradiation swelling and creep, strength changes due to matrix hardening, helium embrittlement, and surface effects such as sputtering and blistering.The interaction of thermal creep, irradiation creep, and swelling results in complex time, temperature, and neutron fluence dependent stress histories in first wall and blanket structures. These effects reduce compressive thermal stresses during the burn portion of a reactor operating cycle and result in residual tensile stress during the non-burn portion of the cycle. The cyclic nature of these stresses, particularly in a tokamak reactor, and the presence of undetected flaws provide a basis for the application of fracture mechanics. Linear elastic fracture mechanics analysis techniques have been applied to predict component life for several conceptual tokamak fusion reactor designs. These analyses show that the structural life may be limited by growth of initial flaws to a coolant leakage. Results indicate that for neutron wall loadings below 2 to 3 Mw/m², life is likely to be controlled by stresses during the burn period and, at higher wall loadings, by residual stresses during the non-burn period.Fracture toughness properties tend to be reduced by irradiation. Therefore, brittle fracture will be a potentially critical failure mode. Fatigue crack growth and fracture characteristics of the design will affect the operating mode of a reactor and influence the performance of different types of reactors. Tests are currently planned to develop material crack growth and fracture toughness data [1] for candidate alloys because these properties have been shown to be important.     

Fracture mechanics analysis of CT-specimens and vessels using FEM-analysis and simplified solutions

To secure the safety of a structure, a reliable capability of assessing the behaviour of an assumed initial crack is required.In the framework of a research and development program for the analysis of the fracture behaviour of vessels and pipes detailed FE-analyses as well as engineering approaches for instance such as based on the Fracture-Handbook (EPRI/GE) are performed in the Gesellschaft für Reaktorsicherheit (GRS) mbH.To investigate the stable crack growth, numerical models for calculating the J-line-integral (Rice), the J according to the virtual crack extension method (Parks) and a topological model for the crack growth (de Lorenzi) have been implemented to the nonlinear FE-program ADINA.Calculations have been performed on several experimental setups such as fracture mechanics specimens and burst vessels, among others on the experiments CT-36 and BVS 030 of the Materialprüfungsanstalt der Universität Stuttgart. Comparisons between the analytical and the xperimental results showed satisfactory agreement. The one parameter J-integral-concept, based on the comparison of the applied crack driving energy with the crack extension resistance of the material has proven to be a suitable tool to deal with a small amount of stable crack growth until the beginning of crack instability. Both techniques, the FE-analysis as well as the Fracture-Handbook evaluation, have been successfully applied.     

Fracture mechanics analysis of a pressure vessel with a semi-elliptical surface crack using elastic-plastic FEM-calculations

Results of an elastic-plastic three dimensional finite element analysis for a semi-elliptical surface crack inside a pressure vessel is presented. The calculations were performed by the finite element program ADINA, incorporating von Mises yield condition and isotropic hardening. The calculations were performed up to that pressure level where general yield of the ligament takes place. The results of the finite element analysis are compared with figures obtained from analytical procedures of elastic as well as of elastic-plastic fracture mechanics.     

Direct analysis method for probabilistic fracture mechanics

A new method for solving problems of probabilistic fracture mechanics (PFM) is proposed. A process of crack growth is reduced into an iterative integration equation with respect to the probabilistic distribution functions of crack geometry using approximate independence, which we have introduced. The integration equation which has a form of Stieltjes integral is solved by a numerical method. Some PFM problems are solved using the present method, and the results are compared with those by the MC method. Failure probabilities obtained from both calculations agree well. Execution time of the present method is shown to be remarkably short.     

Fracture mechanics evaluations of neutron irradiated Type 321 austenitic steel

The effects of fast reactor irradiation at temperatures of ~ 230° C and ~ 400° C on the fracture toughness and associated strength changes, induced in solution treated Type 321 stainless steel have been characterised using instrumented impact test procedures. The studies cover irradiation exposures in the range 16 to 43 displacements per atom (dpa) and test temperatures of 23–500° C.Irradiation results in significant but not catastrophic reductions in fracture toughness, together with radiation hardening effects. Both the dose and test temperature dependence of the toughness changes are sensitive to irradiation temperature. Thus, whilst maximum toughness loss occurs at or below 16 dpa for both irradiation temperatures, the 400° C-irradiation condition is associated with subsequent saturation of the toughness change, whereas for 230° C-irradiation measurable but low on-going toughness degradation occurs up to 43 dpa. The fracture toughness characteristics correlate with fractographic observations which demonstrate retention of a predominantly ductile fracture mode after irradiation, but with dramatic refinement in the scale of microvoid coalescence associated with TiC precipitates. It is suggested that the fracture mechanism after irradiation is controlled primarily by the irradiation-induced precipitate distribution, and furthermore, that the maintenance of ductile fracture, and hence good toughness, up to high irradiation damage levels is a consequence of inhibition of incipient channel fracture processes by the TiC particles.The application of general yield fracture mechanics to calculate critical defect sizes for unstable fracture in fast reactor wrappers is illustrated. These assessments demonstrate the importance of considering net section yield as an alternative failure criterion in thin section components. Finally, the use of empirical equations as a design philosophy to predict irradiation-induced toughness changes is briefly considered.     

Fracture mechanics evaluation of cast duplex stainless steel after thermal aging

For the primary coolant piping of PWRs in Japan, cast duplex stainless steel, which is excellent in terms of strength, corrosion resistance and weldability, has conventionally been used. Cast duplex stainless steel contains the ferrite phase in the austenite matrix, and thermal aging after long-term service is known to decrease fracture toughness. Therefore, we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secure, even when such through-wall crack length is assumed to be as large as the fatigue crack length grown for a service period of up to 60 years.     

Studies on material properties and structural mechanics of engineered barriers for HLW geological isolation

A massive engineered barrier system (EBS) composed of vitrified waste, carbon steel overpack and buffer material (compacted sodium-bentonite) can be expected to isolate hazardous radionuclides from the human environment.Corrosion, leaching and migration studies of EBS materials have supported the performance assessment of the system. Natural analogue is expected to be a method for the validation of the long-term durability of EBS materials. Geochemistry study of groundwater evolution in EBS supports the site generic performance assessment. Coupled thermo-hydro-mechanical process, mechanical stability and hydrogen gas behavior in EBS are also research items for more realistic evaluation of the EBS.     

Fracture behaviour of tungsten-vanadium and tungsten-tantalum alloys and composites

The influence of both microstructure and chemical composition on the fracture behaviour of tungsten-tantalum, tungsten-vanadium composites and alloys of varying chemical compositions are investigated. Industrial solid solution tungsten-tantalum alloys with different tantalum contents in the as-forged condition are investigated along with different tungsten-tantalum and tungsten-vanadium composites and alloys made by powder consolidation, severe plastic deformation using high pressure torsion and different subsequent heat treatments. To investigate the fracture behaviour, several crack propagation directions in relation to the forging direction and shear direction, respectively, are taken into account. Heat treatment of the composite material results in a more homogeneous distribution of the alloying element and the impacts of these specific heat treatments on microstructure and fracture toughness are discussed. The fracture experiments are performed within a temperature range from room temperature to 600 °C and reveal a strong dependence of the fracture toughness and fracture morphology on temperature and on the microstructure, and hence the processing history of the materials.     

Fracture mechanics analysis of iodine-induced crack growth in Zircaloy-4 tubing between 500 and 700°C

Low ductility failure of zircaloy tubing due to iodine-induced stress corrosion cracking (SCC) can occur up to about 700°C. The time-to-failure behavior of Zircaloy-4 cladding tubes containing iodine has been described by the elastic-plastic fracture mechanics model CEPFRAME for the temperature region 500 to 700°C. The model includes an empirically-determined computation method for the incubation period of crack formation, as a portion of the time-to-failure, as well as an elastic-plastic model for describing crack growth due to iodine-induced SCC. The total life time of the cladding tube is obtained by adding the crack initiation and crack propagation periods. The incubation period is a temperature-dependent function of both the depth of surface damage (both fabrication pits and machined notches) and the applied load, and is 40 to 90% of the time-to-failure. The elastic-plastic crack growth model is a modified version of the stress intensity K_I-concept of linear-elastic fracture mechanics. The extensions of this concept take into account a plastic strain zone ahead of the crack tip, which effectively increases the crack depth, and in addition, a dynamic correction factor for the crack geometry which is essentially a function of the effective crack depth. Unstable crack growth is predicted to occur when the residual cross section reaches plastic instability.Model results show good agreement with experimental data of tube burst tests at 500, 600, and 700°C. The crack growth velocity at all three temperatures is a power function of stress intensity ahead of the crack tip; the exponent is 4.9. The model can estimate time-to-failure of as-received cladding tubes containing iodine within a factor of 2. Application of the model to temperatures below 500°C is possible in principle. Due to the increasing scatter in experimental data, the structural transformation of the cladding by recrystallization, and the growing importance of creep strain, CEPFRAME has an upper temperature limit of approximately 650°C. The model is suitable for use in computer codes describing LWR fuel rod behavior during reactor transients and accidents.     

Fracture mechanics analysis on the initiation and propagation of circumferential and longitudinal cracks in straight pipes and pipe bends

The safety potential against pipe fracture of a nuclear piping system is quantified using as an example the fuel circulating system of the THTR 300 MWe. A comparison of the size of cracks expected to occur during operation in the most unfavourable case with the critical crack sizes from the fracture mechanics aspect is used as a criterium for evaluation.A comprehensive test program was carried out to investigate the dependence of longitudinal and transversal crack sizes upon load (bending and internal pressure), pressure medium and temperature. The theoretical methods for predicting the critical crack sizes were checked by the test results.The conclusion is that pipe rupture does not need to be considered under the conditions investigated.     

Fracture mechanics in creep situation interpretation of results on a test specimen

Fracture mechanics in creep situation is a difficult challenge for the 1990s. In France, CEA Saclay has conducted experimental tests on compact tension (CT) specimens at 650°C in order to investigate crack initiation under creep situations. The constitutive material is the 316SPH austenitic stainless steel used for most LMFR structures.Numerical simulations using SYSTUS code and simplified method analysis were performed on one of the tests (CT specimen at 650°C under constant load) to compare some parameters (notch opening, initiation time) with experimental values. The material constitutive law was represented by the complete elasto-viscoplastic CHABOCHE model for computation. Owing to geometrical characteristics such as thickness, the situation of the CT specimen was likely to be intermediate between plane stress and plane strain assumptions. From C^* parameter, incubation time obtained using the R5 rule was conservative in comparison with the test result.The continuum damage model developed at Ecole des Mines de Paris has also been used to assess creep damage in the notch tip area. The crack initiation time has been deduced from critical damage at characteristic distance (X_c = 0.05 mm). Considering critical damage specifically, for a CT specimen (D_c = 0.05), initiation time obtained was higher than the test result.The results of this study will contribute to the development of a methodology for nocivity analysis of cracks in creep situation.     

Fracture mechanics evaluation of small diameter piping considering the latest experimental results

The results obtained from investigations carried out on austenitic piping of small nominal diameter (DN80 and DN50) are introduced and discussed together with their assessment using fracture mechanics methods. Essential results are summarised as following. The pipes with flaws (fatigue crack) down to a depth to a_max/t=0.51 (DN80) as well as a_max/t=0.62 (DN50) and a circumferential extension of results 2α=120° reached bending angles up to 26°. The ASME collapse load (test collapse load) was exceeded considerably and the experimental maximum load could not be reached. Failure due to a leakage or rupture did not occur in any test. The maximum crack extension was 0.69 mm (DN80, a_max/t=0.51) resp. 0.3 mm (DN50, a_max/t=0.62). The experimental maximum load can approximately be assessed by the limit analysis. The fracture mechanics approximation methods GE/EPRI and LBB/NRC calculated a/t=0.4 and 2α=120° initiation loads above the experimental maximum load for pipes containing flaws. These results confirmed the procedures for the proof of integrity of small diameter piping by updating information on load, deformation and failure behaviour of austenitic piping damaged with circumferential flaws. Using these results may formulate a final safety concept for the proof of integrity of small diameter piping by completing the current concepts.     

Fracture mechanics investigations on cylindrical large scale specimens under thermal shock loading

To investigate the crack growth and crack arrest behaviour of primary circuit materials large scale experiments were conducted on component-like specimens under pressurized thermal shock loading at MPA Stuttgart. The material characteristics varied from high tough material to low tough material with higher nil ductility transition temperature to simulate EOL or beyond EOL-state. All tests started from in-service conditions and were cooled down to room temperature. The specimens showed both stable and unstable crack growth and partly crack arrest. The crack growth behaviour was verified by post test calculations and could be explained with the help of the multiaxiality of the stress state.     

A thin-layer reference material for hydrogen analysis

In many areas of material sciences, hydrogen analysis is of particular importance. For example, hydrogen is most abundant as impurity in thin film materials - depending on the deposition process - and has great influence on the chemical, physical and electrical properties of many materials. Existing bulk reference materials (RMs) are not suited for surface sensitive analytical methods like elastic recoil detection analysis (ERDA) or nuclear reaction analysis (NRA). To overcome this serious lack of (certified) thin-layer reference materials for the determination of hydrogen in the near-surface region (1-2 μm depth), we produced stable, homogeneous amorphous silicon layers on Si-wafers (aSi:H-Si) by means of chemical vapour deposition (CVD), while about 10% of hydrogen was incorporated in the Si-layer. Homogeneity and stability were proved by NRA whereas traceability of reference values has been assured by an international interlaboratory comparison.     

Life extension simulation of aged reactor pressure vessel material using probabilistic fracture mechanics analysis on a massively parallel computer

This paper describes a probabilistic fracture mechanics (PFM) computer program using the parallel Monte Carlo (MC) algorithm. In the stratified MC algorithm, a sampling space of probabilistic variables such as fracture toughness value, the depth and aspect ratio of an initial semi-elliptical surface crack is divided into a number of small cells. Fatigue crack growth simulations and failure judgements of those samples are performed cell by cell in parallel. The developed PFM program is implemented on a massively parallel computer composed of 512 processors. As an example, some life extension simulations of aged reactor pressure vessel material are performed, taking analysis conditions of normal and upset operations of PWRs. The results show that cumulative breakage probabilities of the analyzed model are of an order of 10⁻⁷ (1/crack), and that parallel performance always exceeds 90% owing to an employed function of dynamic workload balancing. It is also demonstrated that the degradation of fracture toughness values due to neutron irradiation and the probabilistic variation of fracture toughness values significantly influence failure probabilities.     

Threat analysis for transport of radioactive material using morphological analysis

Abstract

The Swedish Defence Research Agency (FOI) was commissioned by the Swedish Nuclear Power Inspectorate to carry out a pilot study which would serve as the basis for a revised set of regulations regarding physical protection and administrative routines for the transport of radioactive material. The pilot study was to develop a prototype model by which a comprehensive threat analysis could be carried out. The study employed computer aided morphological analysis, which is a flexible, non-quantified modelling method developed at FOI during the 1990s. The paper will present the methodological foundations of morphological analysis and present the prototype models involving general threat scenarios, transport situations, antagonists and strategic measures.     

核磁共振岩心实验分析的标准样品研究

按照国家一级标准的规范,制作了核磁共振（NMR）岩心实验分析的系列化标准样品一流体标样、离散体系标样和固体陶瓷标样,通过对这些标样以及天然岩心的实验考察,将其应用于大庆、新疆、大港等油田的NMR岩心实验分析。结果表明,由于流体标样、固体陶瓷标样和天然岩心的弛豫机理不同,其对岩心NMR孔隙度的标定效果不同。流体标样适合于贝瑞砂岩、陆相沉积分选好、泥质含量低的砂岩NMR孔隙度的标定;固体陶瓷标样适合于分选中等的泥质砂岩NMR孔隙度的标定;对粘土含量高的砂岩、含顺磁性物质的砂岩和砾岩等复杂岩性岩样,用流体标样和固体陶瓷标样标定都得不到准确的NMR孔隙度值。为此,建议选用本地区有代表性的天然岩心作为标样标定其孔隙度,或者研究该类岩石内部磁场梯度分布,得到经内部磁场梯度校正后的NMR孔隙度。     

Fundamental study on applications of damage mechanics to evaluation of material degradation

Damage mechanics theory based on continuum mechanics has recently been attracting attention. This branch of physics describes the mechanical behavior of materials damaged through the production of microscopic voids. Nuclear structural components during operation seldom generate such voids. Damage mechanical theory may be of use for evaluating microscopic material behavior, and confirmation of this point was the purpose of this study. A comparison was made of calculation results, using constitutive equations of Chow and Wang, and of Lemaitre to describe ductile behavior of notched tension bars. The results were also compared with experimental results.