[en] This report is the first technical report to arise from the PISC III project, and it is to be noted that there are some differences from the several reports issued in the PISC I and PISC II projects. The programme on which this report is based, the Centrifugally Cast stainless steel Round Robin Test (CCSSRRT), was designed to identify for further testing those procedures which showed the best promise for effective detection of thermal fatigue cracks while also effectively classifying correctly any innocuous indications. Part of the NDE testing plan, destructive examination, and data evaluation phase were commenced or planned before this programme was incorporated into the PISC III programme, and followed the pattern of the preceding Piping Inspection Round Robin (PIRR). Each of these phases was conducted in a somewhat different way from the procedures adopted in the PISC I and PISC II exercises. However, because the PISC III Management Board considered that the early completion of this work was essential to the establishment of the wider PISC III stainless steel programme, it should be completed as planned, and the results reported because of their general interest and importance. This was done even though the nature of the programme meant that the performance values reported have wide confidence limits. (author)

[en] Pressure boundary components in operating nuclear power plants are subjected to a variety of loading and environmental conditions which could lead to fatigue and stress corrosion cracking. On-line monitoring provides the means of tracking these damage mechanisms and evaluating structural margins during plant operation. This in turn, allows the plant operator to take immediate corrective actions and/or plan more effectively for improvements. Two types of monitoring schemes -material monitoring for cracks and fatigue monitoring - are described. A specific case history of plant application and the resulting benefits of on-line monitoring are also described. (author)

[en] The execution and results of gas pressure tests made on six steam generators for the THTR 300 nuclear power reactor at Schmehausen, West Germany is described. The gas pressure test provided a simple and economic method for detection of major leaks with higher accuracy, cleaner test conditions and less risk of subsequent corrosion than would have been the case if they had made an hydraulic test. (author)

[en] In this paper, the burst pressure of the containment of a pressurized water reactor after a loss-of-coolant accident is calculated on the basis of fracture mechanics principles. The failure criteria applied include linear-elastic fracture mechanics, plastic instability and the two-criteria method. Experiments with plates containing semi-elliptical surface cracks have been performed to assess these methods for 15 MnNi63 containment steel. A leak-before-break evaluation showed that cracks with depths of less than a quarter of the thickness can be expected to exhibit unstable fracture of the spherical part of the containment. (author)

[en] A computer program is developed for the design of pressure vessels. The design rules of the ASME Boiler and Pressure Vessel Code Section VIII Division 1 are applied. The program is written in a graphical utility version of GWBASIC, which has a wide variety of uses among IBM and/or IBM compatible PC users. The program has a wide range of choices for the selection of materials which have been specified by the Design Code. (author)

[en] Three kinds of A508 Cl.3 forging for light water nuclear reactor pressure vessels are used to study dynamic strength and dynamic fracture toughness. The dynamic yield and tensile strengths are described as functions of temperature and strain rate. The plane strain dynamic fracture toughness is most appropriately measured by CT test using a high speed, hydraulic servo-controlled tensile testing machine. The 10 mm thick specimens which do not satisfy the plane strain conditions presented by ASTM E399 can give dynamic fracture toughness values which are small as plane strain dynamic fracture toughness obtained with 100 mm thick specimens only of the condition of Psub(max)/Psub(Q) ≤ 1.1 is satisfied. The plane strain dynamic fracture toughness is not affected by the rate of increase in stress intensity factor over its wide range. Its minimum value is well related to 2 mm V-notched Charpy impact energy. (author)

[en] There is a hierarchy of design code requirements for pressurised components, starting with non-nuclear codes as the minimum and progressing through the ASME III nuclear Classes 3, 2, 1. In establishing and assessing the safety justifications of nuclear plants it is important to have an appreciation of the gradation of requirements in the ASME III design rules and how these go beyond non-nuclear component design rules. There are two broad aspects to the structural integrity of pressurised components, namely the achievement of integrity and the demonstration of integrity. The technical requirements of design codes are associated with achieving integrity while the documentary aspects are usually associated with demonstrating integrity. In practice documents also have a part in achieving integrity in the communication of information between different organisations and personnel involved in the design process. It is not possible to assign simple numerical measures to the relative integrity afforded by non-nuclear codes and the three Classes of ASME III. Instead it is necessary to compare the different requirements of the rules for the various technical and documentary aspects. This paper summarises the most important technical and documentary aspects of the three Classes of the ASME III Code for vessels and the non-nuclear code BS 5500. A similar summary is also provided for the three Classes of ASME III rules for piping. The intention is that the paper provides a basis for appreciating the relative integrity afforded by these various rules. (author)

[en] The reactor pressure vessel has been repeatedly cited as a primary concern in assessment of pressure boundary structural integrity and in planning for plant life extension programs. The life of the reactor pressure vessel will be limited by radiation-induced embrittlement; this is monitored in Westinghouse designed nuclear steam supply systems by testing samples of base metal, heat-effect-zone and weld metal in the form of Charpy V-notch, tensile, and fracture mechanics specimens which have been irradiated in surveillance capsules adjacent to the wall. As data became available from power reactor surveillance and test reactor programs, estimates of radiation-induced changes in mechanical properties were predicted in the form of radiation damage trend curves which provide methods for calculating numerical estimates of changes in mechanical properties as a function of chemistry and fluence. Automatic submerged arc welding was employed in the fabrication of reactor vessels in Westinghouse designed nuclear steam supply systems. The type of flux material utilized in the welding process is important because mechanical properties can differ depending upon what flux is used. This paper correlates the results from over 50 surveillance capsules with the welding practice and concludes that radiation damage trend curves can be developed for welding practice. By using trend curves based on welding practices, discrepancies in chemical analyses are eliminated and credibility is restored to structural integrity assessments. (author)

[en] This paper presents a fracture mechanics evaluation of a reactor vessel nearing the end of its life. The anlysis procedure used is based on a new method for calculating K₁ at crack arrest, and was developed from crack arrest experiments performed by Combustion Engineering for EPRI. The method, already used to analyze data from large specimen crack arrest tests, demonstrates that cracks initiated during severe pressurized thermal shock (PTS) conditions will propagate very much shorter distances than previous methods indicated. (author)

[en] Impurity effects are discussed from the viewpoint of intergranular stress corrosion cracking (IGSCC) of sensitized austenitic stainless steels in simulated reactor water. By using constant elongation rate tensile (CERT) testing it is demonstrated that low concentrations of certain, commonly encountered impurities may have a large effect on the tendency for IGSCC. This is especially true for sulfate and chloride ions and also, but to a smaller degree, for nitrate and carbonate. The effect of chloride is enhanced by copper ions and by hydrogen peroxide in start-up environment. (author)