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Research projects on ageing management

  • As other industrial plants, nuclear power plants are subject to ageing.
  • To prevent that ageing effects have impacts on the safety of German nuclear power plants, the operators carry out certain measures. The entirety of these measures is referred to as ageing management.
  • The BfE co-ordinates research projects with the objective to better understand ageing mechanisms in nuclear power plants and to effectively control or minimise possible effects.

All German nuclear power plants are in operation for over 20 years. As all industrial installations, they are subject to ageing. This means that the original characteristics of the plant components may change, either over time or as a result of operation. Causes for this are various mechanisms such as

  • embrittlement,
  • fatigue,
  • corrosion or
  • wear and tear.

Ageing management

Ageing effects may have an impact on the safety of a nuclear power plant. To prevent this [from happening], the operator carries out certain measures. The entirety of these measures is referred to as ageing management.

Ageing management includes the following measures

  • monitoring,
  • periodic inspections,
  • maintenance and repair,
  • replacement of plant components.

The competent federal state authorities supervise the ageing management at nuclear power plants for technical installations that are important from the safety point of view and have been determined previously.

Research projects on ageing management

Despite of the planned phase-out of nuclear energy by 2022, ageing management remains a central component of the safety philosophy of German nuclear power plants. To ensure this and to answer open questions, the Federal Ministry for the Environment, Nature Conservation, Building and Nuclear Safety (BMUB) initiates studies in this field. The results are to contribute to a better understanding of ageing mechanisms and to control or, respectively, minimise their possible consequences.

Research projects currently co-ordinated by the BfE

Studies on the reliability of fuel elements and mechanical equipment in nuclear power plants

In the context of the research project, it is planned to evaluate systematically the operational experience with fuel elements, pressure-retaining components and mechanical installations such as supporting structures. In this area, various types of damage have occurred in national and foreign facilities. The current state of knowledge is to be updated systematically regarding the causes having led to these damages relating to design, manufacture and maintenance. Conclusions are to be derived for German plants with regard to the reliability of the installations, the effectiveness of measures and the requirements in the German rules and regulations.

Selected concluded research projects co-ordinated by the BfS

Central investigation and evaluation of manufacturing defects and operation damages with regard to pressure-retaining components of nuclear power plants – lot 1 (1 August 2013 until 30 September 2015)show / hide

The objective of the research project was to examine and to evaluate manufacturing defects and operating damages with respect to the pressure-retaining components of nuclear power plants. For this purpose, the international state of the art of safety verifications for break preclusion or, respectively, leak before break, was processed. Pressure-retaining components complying with the requirements of break preclusion or, respectively, leak before break, do not show catastrophic break behaviour in case of overload. A leak forming in case of overload would remain stable to a very large extent. This would provide sufficient time to detect the overload of the component and to respond accordingly, before complete failure would have to be expected. The safety verification for break preclusion and leak before break has so far been carried out deterministically, but probabilistic verifications are increasingly applied. A leak before break verification on a purely probabilistic basis has not been carried out so far. The procedure used in Germany corresponds to the international state of the art of science and technology.

Furthermore, the influence of the medium in determining the fatigue of reactor pressure vessel internals was evaluated. According to present knowledge, significant influence on the damage through fatigue is not to be assumed, these issues continue to be the subject of current research.

The objective of the research project was to examine the performance of ultrasonic tests regarding the detection of errors in austenitic and dissimilar weld seams from special, non-corrosive chromium-nickel types of steel and of quasi-laminar reflectors. One set of test bodies with different weld seams was examined. Differences in the grain structure change the signal patterns only slightly, while the work on the root area has larger influence; however, it was possible to detect all errors. The weld seam structure influenced the signal pattern of the errors but it was possible to detect all errors from 5 mm on. A large container, which is known to have quasi-laminar reflectors, was examined with ultrasonic testing used at present in the German plants. The reflectors were reliably found.

Furthermore, the fatigue resistance of austenitic cladding was examined. It was shown that the fatigue curves for austenitic materials stated in the current KTA safety standards may also be used for austenitic claddings in the area under examination.

Central investigation and evaluation of manufacturing defects and operation damages with regard to pressure-retaining components of nuclear power plants – lot 2 (1 December 2013 until 31 December 2015)show / hide

The objective of the research was to conduct a literature research for three different topics and to update and evaluate the state of knowledge. The initial focus was on primary water stress corrosion cracking. According to present knowledge, the mechanism has not been understood completely. Due to the material concept of German plants, the materials in question were not used frequently, so that the risk of damages based on this mechanism is clearly lower than in foreign plants. In the second place, the impact of earthquake loads, in particular multiple quakes, to material damage (fatigue behaviour in combination with progressive deformation) and the failure behaviour (leakage, break) of pipeline components was analysed. It showed that linear modal analytical earthquake safety assessments carried out according to the response spectrum method generally lead to a conservative evaluation. The third range of topics dealt with strain criteria for structural-mechanical verification procedures in pressure-retaining components. Based on literature research, a limiting strain concept was defined and its applicability to real component geometries and –sizes was demonstrated by numerical analyses and experimental verification.

Central investigation and evaluation of manufacturing defects and operation damages with regard to pressure-retaining components of nuclear power plants (1 April 2012 until 30 September 2013)show / hide

The objective of the research project was to examine the performance of ultrasonic tests with respect to the detection of stress corrosion cracking in dissimilar metal welds. The results enable a first evaluation of the detection capability of the used ultrasonic test parallel to the weld in dissimilar metal welds. The standard ultrasonic inspection technology is able to detect realistic test errors with sufficient signal-to-noise ratio.

Investigations of the effectiveness of measures to ensure the integrity of pressure-retaining components in German nuclear power plantsshow / hide

In the scope of the research project the state of knowledge on the break preclusion con-cept was updated. Furthermore, the procedure for the safety proof of the break preclusion concept was described and the area of application in German plants was analysed. Subsequently the German application of the concept was compared with the US approach. It has shown that the development of the break preclusion concept is widely reflected in the statements and guidelines of the Reactor Safety Commission (RSK). Regarding the works on the monitoring concepts it was stated that they enjoy a high level of development in German plants.

The final report can be found on the GRS website (only in German).

Damage-mechanical modelling of the residual load-bearing capacity of damaged steam generator tubesshow / hide

In the research project, the load-bearing capacity of damaged steam generator tubes was determined at original steam generator tube material with the help of a newly developed damage-mechanical model. In the scope of the project it was determined that the deformation and failure behaviour of the examined steam generator tubes can be described with good accuracy by the simulation.

Experimental investigations to determine the fatigue-resistance of the austenitic cladding of nuclear componentsshow / hide

The objective of the investigation was the experimental identification and validation of a fa-tigue characteristic for the cladding material and the classification of the results in the data base containing the materials used in German nuclear power plants. Key result of this project was the finding that the fatigue curve for austenitic materials according to the rules and regulations of the Nuclear Safety Standards Committee in the examined range of the amplitudes can also be used for the evaluation of the fatigue behaviour of austenitic claddings.

A short summary of this topic can be found on pages 30 and 31 of the Annual Report 2011 of the Materialprüfungsanstalt Stuttgart (only in German).

Evaluation of the effectiveness of ageing management of technical systems in German nuclear power plantsshow / hide

The objective of the research project was to further develop methodically the technical decision-making bases for the Germany-wide evaluation of the effectiveness of ageing management Additionally, a current plant-generic evaluation of the effectiveness of ageing management was carried out which was implemented in German nuclear power plants for safety-relevant technical systems.

The final report on this project can be found on the GRS website (only in German).

You can obtain final reports either via the researchers' websites or upon request from the BfE.

Note: Since 30 July 2016, the Federal Office for the Safety of Nuclear Waste Management (BfE) has supervised projects on ageing management. It took over this task on 30 July 2016 from the Federal Office for Radiation Protection (BfS) that had been responsible until then.

State of 2017.07.10

© Federal Office for the Safety of Nuclear Waste Management