Quality status of components
As part of the quality status of a nuclear power plant, the quality of its safely relevant systems, structures and components (SSC) has to be assessed. The requirements for SSC are determined during the planning stages (design) and implemented using specifications for design, construction and inspection. The same applies to repairs and retrofitting activities which may well be carried out subsequently.
The technical documentation of the SSC, the verification management of the evaluations and adherence to the specifications for repeated inspections are usually distributed in various departments in many different files. With our concept of having a self-contained representation of the SSC quality status, we systematically summarise the actual state including all documentation and statements. This means that changes to requirements and regulations are taken into account in just the same way as for modifications to the components or advances in technical and scientific knowledge.
This approach can significantly reduce the time needed for discussions with assessors and authorities, as the answers to detailed questions can be supplied quickly. Particularly after (unexpected) events, this contributes significantly to keeping downtime to a minimum.
A complete quality status covers the following 6 topics:
- Existing design
- Relevant loading
- Stress analysis
- Fatigue analysis
- Fracture mechanics analysis
- Operating surveillance
1. Existing design
The description of the existing design provides an actual overview of the component. Of particular importance is a description of the design characteristics. Furthermore, the materials used and the welding process are presented. The quality assurance activities accompanying the specifications used at the time of installation or repair/retrofitting are compared with current specifications.
The periodic inspections carried out on the components until now are also presented. Inspection results are assessed according to the requirements of current regulations.
The description of the existing design is the basis for the following steps.
2. Relevant loading
In the step "relevant loading", the real load history as well as the specified loads are discussed.
The load history is often different from the loading anticipated in the design phase. Using measurements taken during operation and/or additional monitoring, realistic amplitudes and gradients can be determined.
Furthermore, potential loading due to specified load cases are summarised. This is used to determine a load envelope for the individual loading stages. This also includes the references where special loads such as pressure surges and earthquake loading have been identified.
3. Stress analysis
The stress analysis shows a summary of the results of existing calculations and, if necessary, supplementary calculations.
- design dimensioning calculations
- subsequent dimensioning calculations according to nuclear safety standards
- pipe calculations
- stability verifications
- flange calculations according to nuclear safety standards or DIN EN and
- finite element calculations
are used to evaluate and assess the stresses.
Furthermore, the required information is provided for
- fatigue analysis
- fracture mechanics analysis and
- future operating surveillance
- sectional loads and
are extracted for further analysis.
4. Fatigue analysis
Nuclear safety standards require fatigue analysis to be carried out – depending on the type of component. Geometrical proportions, material characteristics and pressure ranges, temperature differences and load ranges to be safeguarded according to amplitude and number of cycles are known on the basis of the evaluations of the previous workflow.
This information provides the most important parameters for fatigue analysis. Generally, the simplified procedures specified in the regulations are sufficient for assessment.
Additional investigations and assessment may be required for components which combine media with different operating temperatures. Thanks to our experience with measurements and calculations , we are familiar with these kinds of tasks.
Furthermore, in recent years the issue of environmental assisted fatigue (EAF) has risen. Based on measured data from operating surveillance, we are able to determine realistic correction factors according to NUREG CR-6909.
5. Fracture mechanics analysis
In fracture mechanics analysis, it is investigated whether – within the scope of verification – it is possible for a crack to initiate due to operating loads and loads resulting from faulted conditions. Furthermore, it is investigated whether failure of a component with a postulated crack can be ruled out. It is also checked whether crack propagation can be assumed, and if so, whether crack propagation affecting the integrity of the component is possible between two non-destructive inspections.
We use calculation programs (including the software XPipe to apply the R6 method) and comprehensive fracture mechanics characteristic data for these investigations.
6. Operating surveillance
All the information required to assess the components is available from the previous steps. Therefore, it can be checked whether the previous monitoring activities were sufficient, or if further measures need to be taken.
The monitoring activities can essentially be divided into:
- Monitoring the possible causes and
- Inspecting the consequences of potential damage mechanisms
To safeguard operation, monitoring the possible causes of operational damage has the highest priority. In order to have accurate knowledge of the operational loading, this should be monitored using measurements taken during operation.
Inspecting the consequences of potential damage mechanisms is essentially performed using non-destructive testing. The areas to be monitored and the inspection cycles can be specified based on evaluations from stress and fatigue analysis, combined with current knowledge and the requirements laid down by the regulations.