Papers by Keyword: Reactor Pressure Vessel

Abstract: When the reactor pressure vessel (RPV) is subjected to pressurized thermal shock (PTS), the cooling water injected by the emergency core cooling system (ECCS) will generate a large temperature difference in the wall thickness of the pressure vessel. On the other hand, the fracture toughness of the RPV material decreases a lot under long-term neutron irradiation. Under this condition, the PTS transient may cause a rapid growth of defects in the inner surface of the vessel, resulting in failure of the pressure vessel. In this paper, the fracture mechanics analysis method of RPV under pressurized thermal shock is studied. The thermal analysis and structural analysis of the pressure vessel are performed by finite element method. The stress intensity factor and fracture toughness are obtained through calculation. At the same time, the influence factors of fracture mechanics analysis of RPV under PTS condition are analyzed. The effects of different crack size, crack type, load transient, and neutron irradiation flux on the PTS fracture mechanics analysis results are evaluated. Results show that the larger the ratio of length to depth for axial inner surface cracks, the easier RPV crack grows. Under small break condition, the circumferential cracks are safer than axial cracks. The longer the operating time, the more severe the embrittlement of RPV materials, which will lead to the failure of RPV more easily. For the two typical PTS transients studied in this paper, the re-pressurization condition is safer than the small break condition. The results can provide basis for structural integrity assessment of RPV under PTS condition.

Abstract: During a long-term operation of nuclear power plants (NPP), the changes of structural material properties occur. To ensure the safe and reliable operation, it is necessary to monitor and evaluate these changes mainly on components from primary circuit of NPPs. One of the dominant ageing mechanisms of NPP components besides the radiation embrittlement and the fatigue loads is the thermal ageing. The thermal ageing is the temperature, material and time dependent degradation mechanisms due to long-term exposure at the operating temperature of 570 K.This paper describes the project for thermal ageing monitoring at primary piping in NPP Bohunice Unit 3. There are summarized the results obtained from evaluation of original primary piping material.

Abstract: The elasto-plastic analysis for Reactor Pressure Vessel seal is necessary because of nuclear safety is actively demanded. The existed work based on simplified way to simulate the seal ring, such as uniform stress, spring elements, is too conservative. Therefore, it is necessary to simulate seal ring using 3D solid model considering the elasto-plastc deformation. In this paper, the two-dimensional model is adopted to simulate the compressing and springback character of inconel 718 O-ring using nonlinear isotropic hardening constitutive model and contact algorithm. The simulation approach and constitutive model are validated according to the deformation of seal ring using ANSYS. The seal analysis for reactor pressure vessel is achieved by considering complex loads, such as bolt pre-tightening force and temperature/pressure transient with the usage of an elasto-plastic constitutive model. It is found that the deformation of seal ring obtained by finite element is close to true value. The application of 3D solid model can reduce excessive conservatism effectively and improve the precision of seal analysis compare to the simplified method.

Abstract: The present work explores the importance of model parameters and input variables when simulating the quenching of thick sectioned nuclear forgings. The modelling approach adopted uses values of specific heat capacity, containing latent heat release, to simulate cooling curves; rather than calculating transformation kinetics based upon a mathematical model. Termed the effective specific heat (Cp_eff), two different methods were used to establish values: differential scanning calorimetry (DSC) and thermos dynamic predictive software. Values were then included in finite element (FE) models to simulate the characteristic cooling at the mid-wall position in a thick section forging and were validated against production thermocouple data. The investigation found that the formation of ferrite, bainite and martensite or lower bainite were all represented by the data established using DSC and critical formation temperatures were comparable with others in the literature. Conversely, values calculated using the thermodynamic software failed to represent ferrite formation and predicted different critical transformation temperatures for bainite. The simulated cooling curve that used the software predicted Cp_eff data was comparable to the thermocouple data either side of the bainite transformation, however during the transformation the effects of latent heat on cooling rate were over predicting leading to disparities. The equivalent DSC cooling curves produced a near exact match.

Abstract: The A508-III steel is widely used to manufacture the lower heads of commercial reactor pressure vessels (RPV). In severe accident, the reactor core in the RPV begins to melt and meanwhile the technology of in-vessel retention (IVR) exerts its role. In this case the inner surface of RPV will expose to temperatures over a phase transition temperature. However, the significant nonlinear feature of creep curve of A508-III steel suffered heterogeneous damage was not studied. In this work, the creep tests were performed for the steel at the phase transition temperature of 800°C. The microstructural evolution at different creep stages was characterized by scanning electron microscopy and transmission electron microscopy. The results show that, at the second creep stage, more coarsening second phase particles occur in the steel. With the creep processing, the grain size and diameter of second phase particles increase. At the tertiary creep stage, the grain size increases significantly, and the second phase particles coarsen during the process of atom migration. In addition, Micro-cracks and voids also come into being in the situation and they can become larger by combing each other during the creep process. At this stage, the growth of cavities and second phase particles coarsening become the main mechanism of creep damage. The trend of microstructural evolution is consistent with the creep constitutive equation obtained for the A508-III steel at the phase transition temperature of 800°C. The results obtained provide indispensable foundation to establish the relationship between the macroscopic creep and microscopic damage.

Abstract: Most of the French Nuclear Power Plants (NPPs) are currently embarking upon efforts to renew their operating license, while the pressurized thermal shock (PTS) events and environmentally assisted fatigue (EAF) pose potentially significant challenges to the structural integrity of the reactor pressure vessel (RPV) which has the potential to be NPP life-limiting conditions. In the assessment of the PTS events, the deterministic fracture mechanics (DFM) is still used as the basic mechanics in most countries except for the USA. While the maximum nil-ductility-transition temperature (RT_NDT) is about 80°C for 54 French RPVs after 40 years operation, the maximum allowable RT_NDT is only about 70 ^oC and 80 ^oC for the typical PTS events in the IAEA and NEA reports, respectively. On the other hand, the effects of light water reactor (LWR) environmental (other than moderate environment in the code) were not considered in the original design, while the effects of LWR environmental are needed to be considered in the LRA according to the USA regulations. In this paper, the challenges of the PTS and EAF are discussed, and some suggestions are also given for the LRA

Abstract: One potential challenge to the integrity of the reactor pressure vessel (RPV) in a pressurized water reactor is posed by pressurized thermal shock (PTS). Therefore, the safety of the RPV with regard to neutron embrittlement has to be analyzed. In this paper, the procedure and method for the structural integrity analysis of RPV subjected to PTS is presented. The FAVOR code is applied to calculate the probabilities for crack initiation and failure by considering crack distributions based on cracks observed in the Shoreham and PVRUF RPVs in the U.S. A local approach to fracture, i.e. the σ*-A* model is used to predict the warm prestressing (WPS) effect on the RPV integrity. The results show that the remaining stress contributes to the WPS effect, whereas the increase of fracture toughness is not completely attributed to the remaining stress. The modeled load paths predict a material toughness increase of 30-100%.

Abstract: Service life of VVER-type nuclear reactor is limited by decrease in brittle fracture resistance of reactor pressure vessel produced of low-alloy low-carbon steel under effect of irradiation and/or elevated temperatures. In this work fracture surfaces were studied by Auger-electron spectroscopy in order to estimate the contribution of intergranular embrittlement to the degradation of reactor pressure vessel steels under the influence of operating conditions. It was demonstrated that irradiation induced segregation leads to an increase of P content in grain boundaries that promotes intergranular brittle fracture on fracture surfaces. The similar effect but to a lesser degree was shown in the case of long-term temperature exposure. The grain boundary structure was examined and an effect of carbides located on the grain boundaries is supposed due to increased phosphorus segregation on carbide/matrix interface boundaries.

Abstract: Reactor Pressure Vessel is the core equipment in the nuclear power plant, many problems happen during the manufacturing process of localization. Through the statistical analysis on non-conformance and problems detected by quality surveillance, the characteristics and difficulties of RPV quality control in domestic manufacturing process is researched, and the basic problem and the direction of improvement has been analyzed. The data, contents and opinions will further provide reference for RPV and other similar equipments manufacturing quality control.

Abstract: The composition and heat treatment of heavy section forging steel for reactor pressure vessel were optimized applying physical and numerical simulation methods, including numerical simulation to calculate temperature distribution during quench of model forging, artificial neural network to predict CCT diagrams of steels and small sample control cooling to simulate specific heat treatment. And the influence of compositon and heat treatment on microstructure and properties were discussed. Results showed that the experimental steel obtained satisfactory properties based on optimization of chemical composition and heat treatment. It is estimated that the hardenability and temper stability of experimental steel were improved by tungsten alloyment and higher temperature temper was good for superior microstructure, proper strength and better toughness. In the present work, application of simulation methods is proved to be reasonable for study on heavy section forging steel.