The French Nuclear Safety Authority (ASN) considers that the serviceability of 900 MWe reactor pressure vessels is demonstrated for up to 40 years

Published on 05/11/2010 • 05:00 pm

Information notice

In order to take into account pressure vessel aging and possible changes in available knowledge and information, the demonstration of the serviceability of NPP pressure vessels is regularly updated by EDF and examined by the French Nuclear Safety Authority (ASN). In anticipation of the decisions to be taken regarding the continued operation of each 900 MWe reactor after the third ten-year inspection (VD3), EDF has submitted a justification documents for technical evaluation by the ASN with the technical support of the French Institute for Radiological Protection and Nuclear Safety (IRSN).

After consulting the Standing group for reactor pressure vessel safety, and subject to measures to be taken by EDF in terms of pressure vessel aging monitoring and in-service inspection, the ASN has not identified any generic issues compromising the serviceability of 900 MWe reactor pressure vessels until the next 10-year inspection.

 This generic assessment will be supplemented with an individual analysis of each reactor after the conclusion of the third 10-year inspection.


In-service inspection system during inspection phase

[In-service inspection system during inspection phase]

Pressure vessel serviceability: A key aspect of the safety demonstration process

Pressure vessel integrity is of essential importance for demonstrating the safety of PWR NPPs. This is one of the reasons why measures to ensure pressurise vessel integrity throughout the reactor service life must be implemented as of the design phase. The in-service integrity of pressure vessel components is demonstrated through a mechanical demonstration, an aging monitoring programme and an in-service inspection programme conducted by EDF.

Aging of reactor pressure vessels

The pressure vessel contains the reactor core. It is therefore exposed to high temperatures (300ºC), high pressures and high radiation levels during nuclear power plant operation. The mechanical properties of the pressure vessel steels have been modified for improved radiation resistance. Under the effect of neutron radiation, pressure vessel steels are "embrittled", i.e. their resistance to rupture in the presence of a defect is decreased. EDF has developed a model used to predict the embrittlement of pressure vessel steels during exposure to a given radiation level. This generic model is based on a large quantity of data and is in compliance with practices identified by the IAEA. The theoretical results obtained with this model are supplemented with an experimental verification of mechanical test samples. Specifically positioned inside the reactor pressure vessel so as to be exposed to higher radiation levels than the pressure vessel steel, these samples are regularly extracted and examined to anticipate changes in metallic properties.

Consideration of manufacturing defects

The demonstration of pressure vessel serviceability takes into account aging effects and the possible presence of manufacturing defects, i.e. non-detectable defects with sizes below the detection limits guaranteed by inspection procedures, or defects identified during in-service inspections. Certain pressure vessels of the French NPP fleet exhibit underclad defects [1] due to manufacturing processes. A total of 33 underclad defects have been reported in 9 pressure vessels, including 20 in the Tricastin 1 reactor pressure vessel. These defects are regularly inspected to ensure that they do not evolve during operation, which is currently the case.

Measures to slow down aging and reduce its impact

In order to minimise the embrittlement of pressure vessel steels, EDF has implemented nuclear refueling procedures effectively reducing pressure vessel radiation exposure.

The thermal shock [2] potentially undergone by a pressure vessel under accident conditions can be reduced by heating the safety injection water, for example to ensure the absence of danger due to defects detected during in-service inspection of a section of the pressure vessel subject to significant embrittlement. This measure was already implemented for the Tricastin 1 and Fessenheim 2 reactors during the second 10-year inspection thereof.

Validity of the demonstration of pressure vessel serviceability for the 10 years to come

The ASN and IRSN have jointly evaluated the demonstration of pressure vessel serviceability so as to ensure compliance with regulatory requirements and verify the validity of the calculations and hypotheses used. The objective was to make sure that the results obtained at each step of the calculation are conservative and that regulatory safety margins are met.

The calculations performed by EDF have demonstrated that compliance with regulatory requirements will be ensured throughout the 10-year period after the third 10-year inspection. In the case of the Saint Laurent B1 reactor, the ASN has requested that heating of safety injection water be performed during the third 10-year inspection.

The ASN considers that the serviceability of all 900 MWe reactors has been demonstrated for the 10-year period after the third 10-year inspection. The ASN will verify that the inspections conducted during these 10-year inspections effectively ensure the absence of new defects, as well as the absence of evolution of detected defects. The ASN has noted that EDF is capable of rapidly implementing technical measures such as safety injection water heating whenever necessary to ensure the absence of danger due to defects detected during in-service inspection in the case where new evidence should compromise the present documentation.

The ASN has also formulated several requests aiming to further improve current methods and assess their conservativeness, and to pursue ongoing studies to confirm the data currently available.

[1] Nuclear reactor pressure vessels are covered with a stainless steel cladding to protect the pressure vessel steel against reactor coolant water.

[2] In the event of an accident causing reactor coolant leakage, cold water is injected to cool down the reactor. This massive injection of cold water may cause an abrupt local decrease in pressure vessel temperature. (cold shock).

Date of last update : 08/06/2017