Thermal Ageing Embrittlement of Casting Duplex Stainless Steels for Nuclear Power Plant

Cast duplex stainless steels (CSS) used for PWR pipes are degraded due to thermal ageing embrittlement during long-term service at 288 °C to 327 °C. Z3CN20-09M Cast duplex Stainless Steels (CSS) made in France for domestic nuclear power plants were thermally aged at 400 °C for 100 h, 300 h, 1000 h, 3000 h and 10000 h. The tensile properties and the impact properties at different thermal aging duration were measured and the effects of the thermal aging time on the microscopic structures and substructures of Z3CN20-09M were respectively investigated by optical microscopy and transmission electron microscopy. The results showed that the tensile strengths of Z3CN20-09M CSS increased gradually with the increment of the thermal ageing time, whereas the impact properties decreased with the prolonging of the thermal ageing time. After long thermal ageing time the dislocation configurations were greatly changed in austenite, and there were precipitates along the austenite-ferrite interface. Moreover, the iron-rich α phase and the chromium-rich α phase precipitated in ferrite aged for 10000h by nucleation and growth rather than the spinodal decomposition. All of above revealed that Z3CN20-09M CSS became brittle during thermal ageing.

Experimental Investigation of Copper/Stainless Steel Joints Formed by Vacuum Brazing

The article presents a study of two different brazing joints produced by dissimilar materials vacuum brazing. The junctions were obtained between copper or copper alloy and stainless steel. Different brazing parameters were used according to the different type of samples. By using optical microscope, scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDS) and micro-hardness machine to analyze the microstructure of copper or copper alloy/stainless steel vacuum brazing joins. The test results showed that copper (T2)/stainless steel (1Cr18Ni9Ti) dissimilar materials were successfully bonded together by means of the advanced vacuum brazing technology (the grade of filler metal was B-Ag72Cu). The interface zone of copper (T2)/stainless steel (1Cr18Ni9Ti) brazing bonded joint included the copper side interface, the middle brazing transition zone and stainless steel side. Some defects such as microfissures were also found in the brazing seam between copper alloy and stainless steel composite components obtained by vacuum brazing using B-AgCu21Pd25 filler metal. They are mainly due to the process and geometry parameters, such as temperature and clearance.

The Reliability Analysis of Bearing in Large Blower System

A large blower is one of the important subsystems in some complex gas transport. Typically, the blower system in the gas-cooled reactors act as the main pump in the water cooled reactors. The bearing is an essential equipment of the blower. In the vertical blower system, electromagnetic bearing and auxiliary bearing can be selected to suspend the blower system. Then the auxiliary bearing will bear the weight and kinetic energy of blower in the case of the electromagnetic bearing fails in process of blower running. Therefore, the reliability of the suit of bearing is one of the key questions in the design of blower. According to the design, the reliability of the suit of bearing is depended on the reliability of electromagnetic bearing and the design life of auxiliary bearing which is indicated by the times it can bear blower’s decline in the duration of running. In this paper, the reliability of this kind of bearing will be analyzed through a case study and the risk that it contributes to the blower system will be studied. The method of analysis is based on the reliability distribution of bearing and the Poisson distribution model. The advice about the design of bearing will be given from the results of analysis.

Chemico-Physical Causes of Radiation-Induced “Long Cell” Action Corrosion in Water Cooled Reactors

In the previous overview papers [1, 2], the author has identified that ‘long cell action’ corrosion plays a pivotal role in practically all unresolved corrosion issues, irrespective of reactor types and operation. In trying to confirm the existence of radiation-induced ‘long-cell’ action (macro) corrosion cell in the primary cooling system of LWRs, the author attempted to theoretically reproduce the electrochemical potential difference demonstrated during experiments at the INCA Loop in Sweden and the NRI-Rez Loop in the Czech Republic [3, 4]. By performing a radiation chemistry kinetics study combined with electrochemistry calculations, the hydrated electrons, e−aq, reacting mainly with stable molecules, are found to be responsible for inducing a large portion of the potential difference both in the PWR and BWR water chemistry environment. Considering large uncertainties, the author used the standard equilibrium potential as a fitting parameter in the previous studies [3, 4]. The standard chemical potential of the hydrated electron estimated from the fitting parameter is far less than the generally accepted value of 2.86 V. In order to resolve the large discrepancy between the generally accepted values and the estimation from the fitting parameter, the author has developed a ‘mixed’ radiation-electrochemistry formalism, which enables theoretical reconstruction of the observed potential differences more clearly. The previous verifications are updated by using this approach. Through these studies, the author has confirmed the existence of the ‘long cell’ action corrosion mechanism existing in the water-cooled reactors.

Aspects of Minimizing Steel Corrosion in Liquid Lead-Alloys by Addition of Oxygen

The concept of minimizing steel corrosion in liquid lead-alloys by addition of oxygen strongly depends on the availability of efficient devices for oxygen transfer and reliable oxygen sensors. The accuracy of electrochemical oxygen sensors is analyzed on the basis of theoretical considerations and results from experiments in stagnant lead-bismuth eutectic (LBE). Additionally, the feasibility of gas/liquid oxygen-transfer and the long-term performance of electrochemical sensors in flowing liquid metal are addressed based on operation of the CORRIDA loop, a facility for testing steels in flowing LBE.

Introduction of Hot Isostatically Pressed, Reactor Coolant System Components in PWR Plant

This paper presents an overview of the work undertaken by Rolls-Royce to introduce Hot Isostatically Pressed (HIP) components into Pressurised Water Reactor plant, and also results of non-destructive and destructive examinations of a thick-walled pressure vessel. It presents the work from a design justification/manufacturing quality assurance perspective, rather than from a pure metallurgical perspective. Although the HIP process is not new, it was new in its application to Rolls-Royce designed nuclear reactor plant. As a consequence, Rolls-Royce has implemented an evolving, staged approach, starting with HIP bonding of solid valve seats into small bore valve pressure boundaries. This was followed by powder HIP consolidation of leak-limited, thin-walled toroids, and has culminated in the powder HIP consolidation of thick-walled components. The paper provides an overview of each of these stages and the approach taken with respect to justification. Mechanical testing and metallurgical examination results of sample material taken from different sections of a thick-walled component are presented. A full range of test results is provided covering, as examples: tensile, charpy and sensitization susceptibility. Differences in weldability between the HIPed and the previous forged form are also documented. The paper describes the benefits that Rolls-Royce has realised so far through the introduction of HIPed component. Structural integrity benefits are described, such as improved grain structure, mechanical properties, and ultrasonic inspection. Project-based benefits are also described, such as provision of an alternative strategic sourcing route, cost and lead-time reductions.

Experiences From Structural Dynamic Analysis Projects of BWR Plants Within the Scope of Power Uprate Projects Using FEA

During recent years, power-uprate projects have been executed at several BWR-units in Sweden. As part of these projects, structural verification of the safety-related buildings as well as the new and old internal parts of the reactor pressure vessel, RPV, has been performed. In this document, some experiences will be presented from structural dynamic verification, using finite element analysis, FEA, within the scope of these power uprate projects. From this work, a number of conclusions can be drawn. Global models with dense meshes can successfully be used for a broad range of applications. Today, large FEA-models can be used efficiently, e.g. in global vibration and structural verification analyses, if suitable dynamic analysis methods are used. There can be strong dynamic interactions between the containment, fluids, the RPV and RPV-internals. Stress calculation and evaluation can be executed efficiently on large models. The structural models can with advantage be re-utilized in future projects.

On-Line Sequestration of Fission Product Gases in Operating Nuclear Reactors

A research program is outlined to theoretically and experimentally establish the effectiveness and acceptability of on-line sequestering of fission product gases (FPGs) as they are generated in a fissile fuel pin using adsorbents and/or molecular sieves selectively distributed within the lower spring-plenum region of the LWR fuel. If successful this innovation can be implemented in existing and next generation LWRs and may: • reduce accident source term (releasable FPGs); • enhance reactor safety and acceptance; • extend fuel life suppressing clad swelling and fuel failure; • reduce thermal neutron poisons in high flux core regions; • improve operational and economical performance; • require only minor changes to fuel pin assembly and use; • use in LWRs with no license or tech-specifications impact.

Rupture Frequency of CANDU® Large-Diameter Primary Heat Transport Piping Estimated Using Probabilistic Fracture Mechanics Codes

This paper presents a pilot study of using probabilistic fracture mechanics codes (PRO-LOCA 2009 and WinPRAISE 2007) to estimate the rupture frequency of CANDU® large diameter Primary Heat Transport (PHT) piping. The results of this study show that WinPRAISE 2007 and PRO-LOCA 2009 produce comparable trends for the predicted probability of leak and probability of large break leak. There is a number of sensitive leak detection methods available in CANDU plants. The materials and quality of fabrication and sensitive leak detection results in the total probability of a large break leak in the large diameter PHT piping welds being estimated to be on the order of 1E−8 breaks per plant per year. The results of the pilot study indicate that probabilistic fracture mechanics codes could be used to demonstrate that a shutdown action limit of 100 kg/h is sufficient to ensure the probability of rupture of large diameter PHT piping welds is extremely low.

Investigations on Optimal Design of a U-Tube Steam Generator

Steam generator is one of the key equipments in the pressurized water reactor, from the performance point of view, it is necessary to apply optimization techniques to the design of the steam generator. In this work, the optimal designs of a U-tube steam generator (UTSG), taking minimization of the total volume and net weight as objective respectively, are carried out by considering thermohydraulic and geometric constraints using a complex-genetic algorithm (CGA). And the sensitivities of some parameters which influence the total volume and net weight of UTSG are also analyzed. Under the condition of constant secondary thermalhydraulic parameters of the steam generator, the optimal design indicates an obvious effect taking either the overall volume or the total weight of the steam generator as the objective. The optimization results show that the proposed optimal method is feasible and effective. And the results of optimal designs and sensitivity analysis would provide guidance in the engineering design of UTSG.