M. Erve

Calculation code for erosion corrosion induced wall thinning in piping systems

Abstract Extensive experimental and theoretical investigations have been performed to develop a calculation code for wall thinning due to erosion corrosion in power plant piping systems. The so-called WATHEC code can be applied to single-phase water flow as well as to two-phase water/steam flow. Only input data which are available to the design engineer or the operator of a plant are taken into consideration. Together with a continuously updated erosion corrosion data base containing results from experimental investigations and actual damage in power plants the calculation code forms one element of a weak point analysis for power plant piping systems which can be applied to • - minimize material loss due to erosion corrosion, • - reduce non-destructive testing and curtail monitoring programs for piping systems, • - recommend life-extending measures.

Cracking in stabilized austenitic stainless steel piping of German boiling water reactors—characteristic features and root cause

Cracks have been found in the welds of piping systems made from stabilized austenitic stainless steels in German boiling water reactors (BWR). In the course of the intensive failure analysis metallographic examinations, microstructural investigations by electron microscopy, corrosion experiments and welding tests have been performed. The results show that cracking under the given medium conditions is due to intergranular stress corrosion cracking (IGSCC) in those parts of the heat affected zone (HAZ) which are overheated during welding and where solution of titanium carbides and subsequent precipitation of chromium carbides and depletion of chromium along the affected grain boundaries could occur.

Planned and implemented remedial measures to prevent IGSCC and to ensure safe plant operation with BWR piping systems made of stabilized austenitic steels

After intergranular stress corrosion cracking had been found in piping, made from stabilized austenitic stainless steel, carrying reactor water in German BWRs, remedial measures have been taken with regard to the refurbishing concepts, material considerations, water chemistry and optimized fabrication processes. By taking adequate remedial action the integrity of the pipes for long-term plant operation is ensured. In addition study programs are already under way to provide better knowledge on crack initiation and crack growth behavior.

Environmentally assisted crack growth in low alloy steels — results and their relevance to lwr components

Abstract The paper describes the influence of corrosion on crack initiation and crack growth in low-alloy steels in high temperature water and the relevance of data determined by corrosion tests to the component behaviour. The test results, gained by laboratory experiments and the literature data available are analyzed and evaluated with respect to the in-service conditions of the components (e.g. RPV, feedwater-, main steam-line etc.). As a result of this evaluation, it can be stated that due to the current boundary conditions the RPV and other important components of the primary circuits in the water/steam cycle of LWRs are not endangered by stress corrosion cracking.

State of the art in the manufacture of heavy forgings for reactor components in the Federal Republic of Germany

The aim of system manufacturers and operators, to reduce the number of welds in components and piping to a minimum for technical and economic reasons, has led to an optimised design for the entire reactor coolant system of LWRs as regards forging technology. It is shown, by way of examples, that the technological limits (concerning ingot weights and forging methods) have been continually extended through intensive research and development, as well as through the best use of available manufacturing capacity and equipment. Results from destructive testing and non-destructive examination of component parts are illustrated and evaluated. It is shown how these results and the experience from material qualification have been taken into account in the KTA nuclear safety standards.

Provisions to prevent strain-induced corrosion cracking in ferritic piping systems of boiling water reactors

Abstract Based on failure analysies and evaluations of laboratory test results, system areas are localized which are to be regarded as susceptible to strain-induced corrosion cracking (SICC). Possible systems engineering, operational, design and manufacturing provisions for preventing SICC are described. Necessary measures have already been taken by BWR operators, acting in part together with authorised experts and licensing authorities, in the case of plants which are in operation.

Residual life assessment, technical improvements and backfitting of components and systems from the materials standpoint

Abstract The safety standard which is attained at the time of initial startup by a nuclear power plant built in accordance with state-of-the-art design and engineering principles must be assured throughout the plants entire lifetime. Based on operating experience a plants original design life should be systematically reevaluated in the light of new findings and developments in safety. The results of an analysis of this type can also be useful for the purposes of preventive maintenance or to prolong a plants planned or licensed service life. Moreover, they form the basis of decisions regarding technical upgrades and backfits and are of value in optimizing plant reliability and availability. A concept exists for components in light water reactors which provides for prompt identification and remedying of damage due to the deterioration in service of materials and their properties (e.g. fatigue, local corrosion mechanisms, erosion corrosion, neutron irradiation). It may be necessary or appropriate to implement technical upgrades or backfits where application of more up-to-date safety standards (e.g. break preclusion methodology) demands compliance with more stringent requirements than those originally used in the design basis, or where systematic damage cannot be ruled out (unstabilized austenitic steels in BWR plants), or where it is possible to make substantial reductions in the radiation exposure of operating or maintenance personnel (substitution of cobalt-containing materials) during the plants lifetime.

Activities in the field of plant life evaluation, life extension and plant improvement

Abstract The adequate margin of safety reached in the as-built condition for a nuclear power plant shall be maintained throughout the whole life of the plant. To attain this, a systematic lifetime evaluation of safety related items should be performed in due time in the light of new developments. The results can be used for the purpose of life extension and license renewal, too. Siemens has gained an integral analysis concept by practical experience. The analysis can result in corrective actions with respect to life extension of plant components and systems or can lead to measures for plant improvement. Examples of performed activities are given within this article.

Materials for LWRs in the light of new developments

Abstract One of the pillars of the outstanding safety level and high availability of KWU nuclear power plants is the quality of the materials used. Base material toughness, weld quality and new fabrication methods for piping components are used as examples to describe this high quality. Plastic limit load theory is applied to define the relationship between attainable and specified toughness. The conclusions already drawn and the developments which may conceivably be derived from them are discussed on the basis of the aforementioned facts.

Significance of material technology for plant life management in nuclear power plants

The life-limiting mechanisms for components and systems are physical aging and wear. Both of them are related to changes of microstructure in the bulk material or at the phase boundaries medium/material and material/material. They are triggered during operation by factors such as temperature, mechanical load, and environment. Thus, to achieve an utmost effective aging management it is necessary, to understand the underlying aging and wear mechanisms such as neutron irradiation, fatigue, corrosion, fretting, etc. Definition and qualification of suitable corrective and preventive actions against accelerated aging, requires precise knowledge of the aging processes and life-limiting situations and thresholds. It is obvious, then, that materials engineering plays a large part in effective and economical plant life management. Within this paper, the role of materials science and technology in plant aging management during the various stages within a whole life cycle of a power plant is described: (1) the correct choice of materials as part of a well-based materials concept in the design stage is very important for later plant operation. As an example steam generator materials are presented. (2) The parameters of the individual manufacturing processes during erection of components and systems must be optimally selected in order to guarantee long-term operation. As an example the reasons for core shroud cracking in a BWR NPP are discussed. (3) Aging mechanisms must be accounted for in operation of components and systems, and their effects have to be counteracted in order to prevent service-life limiting situations. Details are described with respect of corrosion and neutron irradiation. Demanding future tasks for materials science and technology are presented, which are necessary to continue to contribute to an optimized plant life management and to cost-effective operation of nuclear power plants at high safety levels.