In this respect, the manufacturer must define and apply the provisions enabling it to justify compliance with these requirements in the design and manufacture of the NPE. As licensee of the reactors, EDF is subsequently responsible for their installation, utilisation and in-service monitoring, in compliance with the conditions set out by the manufacturer and taking account of its own OEF. 2.2.2 Assessment of the design and manufacturing of Nuclear Pressure Equipment ASN assesses the compliance with the essential requirements by the NPEs most important for safety, referred to as “level N1”, corresponding primarily to the reactor pressure vessel, the SGs, the pressuriser, the reactor coolant pumps, the piping, notably that of the Main Primary System (MPS) and Secondary Systems (MSS), as well as the safety valves. ASN can be assisted in this task by organisations that it approves and oversees. The conformity assessment of the other NPE (“level N2 or N3”) is carried out directly by these approved organisations. Four inspection organisations or bodies are currently approved by ASN to assess NPE compliance: Apave Exploitation France, Bureau Veritas Exploitation, Vinçotte International and the inspection body of the EDF users. This conformity assessment concerns the equipment intended for the new nuclear facilities (more than 200 equipment items for the Flamanville EPR reactor) and the spare equipment intended for nuclear facilities already in service (notably the replacement steam generators). It takes the form of an examination of the technical documentation of each equipment item and inspections in the workshops of the manufacturers, as well as at their suppliers and subcontractors. In 2024, an increasing share of ASN’s assessment activities concerned the NPE intended for the EPR 2 reactors. In addition, ASN’s actions were reinforced with respect to monitoring of level N2 or N3 NPE suppliers and manufacturers, focusing more particularly on certain sensitive activity sectors such as foundries or piping manufacturers. The NPE manufacturing operations involve a large number of suppliers, to which the manufacturers sometimes subcontract activities that are decisive for the mechanical strength of the equipment produced. In this context, ASN reinforced its checks, considering that the deficiencies it had identified at certain materials suppliers concern the control of sensitive manufacturing processes (such as welding or casting), or the traceability of manufacturing operations. 2.2.3 Operation of Nuclear Pressure Equipment The reactor MPS and MSS, which contribute to the containment of the radioactive substances, to cooling and to controlling reactivity, operate at high temperature and high pressure. The monitoring of the operation of these systems is regulated by the Order of 10 November 1999 relative to the monitoring of operation of the MPS and the MSS of PWRs. These systems are thus the subject of monitoring and periodic maintenance by EDF. These systems are in particular subject to periodic requalification every ten years, which comprises a complete inspection of the systems involving non-destructive examinations of the most sensitive areas, hydrostatic pressure test and verification of the good condition and good operation of the over-pressure protection accessories. The licensee is also required to keep and update files on the design, manufacture, overpressure protection, materials, findings made during operation and, as applicable, processing of deviations, as often as necessary and at the time of the periodic requalifications. Some of the safety issues of the components of the primary or secondary systems are detailed below. The reactor pressure vessels The reactor pressure vessel is an essential component of a PWR and contains the reactor core and a part of its instrumentation. In normal operating conditions, the vessel is entirely filled with water, at a pressure of 155 bar and a temperature of 300°C. It is made of ferritic steel, with a stainless steel inner liner. Regular inspection of the condition of the vessel is essential for several reasons: ∙The consequences of rupture of this equipment are not studied in the reactor safety case. Monitoring thus contributes to the break preclusion approach adopted for this equipment. This approach The principles of the reactor vessels in-service strength demonstration Very close attention is paid to preventing the risk of fast fracture of the reactor pressure vessel. On the occasion of each periodic safety review, this requires a verification of the strength of the vessels with a view to the next ten years of operation, taking account of the embrittlement of their steel under the effects of irradiation. The risk of fast fracture of a reactor pressure vessel depends on the joint presence of three factors: the presence of a flaw (such as a crack in the metal), which is liable to lead to stress concentration, mechanical loading (as a result of routine reactor operations or an incident or accident situation) and insufficient mechanical strength of the material. The justification process thus relies on analysis of the resistance to fracture of the known flaws in the vessels (which concerns a few EDF reactor vessels, affected by flaws under the liner). In addition, during each ten-yearly outage, the vessels undergo a non-destructive examination of the area subject to irradiation, which is designed to detect flaws of sufficiently large size. To ensure the conservative nature of the demonstrations, the strength of a “reference” flaw, corresponding to the largest flaw that could not be detected during these examinations, must therefore be also checked. For the reference flaw and for the flaws under liner affecting vessels, the analysis of the risk of fast fracture of the reactor vessels thus consists in: • assessing the mechanical stresses liable to lead to initiation of propagation of a crack where these flaws are located, in all reactor operating situations, including an accident; • assessing the characteristics of the material’s mechanical strength, taking account of steel embrittlement under the effect of irradiation. Embrittlement of the steel is assessed every ten years, based on the results of mechanical tests performed on material samples which were placed in capsules inside the reactor cores. The location of these capsules, close to the fuel assemblies, enables these samples to be subjected to higher irradiation than that which reaches the vessel wall and thus provide an early check on the validity of the assessments made. The fast fracture analysis process thus leads to a comparison between the loading at the end of the flaw, which is the result of the first step, with the mechanical strength of the material, which is a result of the second. Safety coefficients are also used in the fast fracture analysis, to ensure that they are sufficiently conservative. Finally, the vessel as a whole undergoes other non-destructive examinations, for example on the closure head, and a pressure test of at least 1.2 times the service pressure. This test is an overall strength test capable of detecting unexpected degradation phenomena. For example, during the test performed in 1991 on the vessel of one of the reactors of the Bugey NPP, a leak was detected from a closure head adapter owing to a stress corrosion phenomenon. This leak led EDF to replace all the closure heads of its 900 and 1,300 MWe reactors between 1994 and 2009. 302 ASN Report on the state of nuclear safety and radiation protection in France in 2024 The EDF Nuclear Power Plants
RkJQdWJsaXNoZXIy NjQ0NzU=