Abstracts ASN Report 2019

the accident situation and in addition to the accident, there will be the most prejudicial failure of one of the components used to manage this situation. As a result of this, the systems brought into play in the event of an accident (safeguard systems ensuring emergency shutdown, injection of cooling water into the reactor, etc.) comprise at least two redundant and independent channels. • nƺɮƺǼ ! يג ȒȇɎȸȒǼ Ȓǔ ƏƬƬǣƳƺȇɎɀ ɯǣɎǝ ƬȒȸƺ ȅƺǼɎƳȒɯȇ These accidents have been considered since the Three Mile ņĿĴŁķ ĴĶĶļķĸŁŇ Ҏӄӌӊӌҏ ĴŁķ ĴŅĸ ŁłŊ ŇĴľĸŁ ļŁŇł ĴĶĶłňŁŇ ļŁ ŇĻĸ design of new reactors such as the EPR. The aim is to preclude such accidents or to design systems that can withstand them. • nƺɮƺǼ يד xǣɎǣǕƏɎǣȒȇ Ȓǔ Ɏǝƺ ȸƏƳǣȒǼȒǕǣƬƏǼ ƬȒȇɀƺȷɖƺȇƬƺɀ Ȓǔ ɀǣǕȇǣǔǣƬƏȇɎ ȸƺǼƺƏɀƺɀ This requires implementation of the measures provided for in the emergency plans, including measures to protect the general public: shelter, taking of stable iodine tablets to saturate the thyroid and avoid fixation of released radioactive iodine, evacuation, restrictions on consumption of water and of agricultural products, etc. ׏ِ א ِ ב ¨ȒɀǣɎǣȒȇǣȇǕ Ȓǔ ƫƏȸȸǣƺȸɀ To limit the risk of releases, several barriers are placed between the radioactive substances and the environment. Barriers must be designed to have a high degree of reliability and must be monitored to detect any weaknesses or failures. There are three such barriers for pressurised water reactors: the fuel cladding, the boundary of the reactor primary system, and the containment (see chapter 10). ׏ِ א ِ ( ג ƺɎƺȸȅǣȇǣɀɎǣƬ ƏȇƳ ȵȸȒƫƏƫǣǼǣɀɎǣƬ ƏȵȵȸȒƏƬǝƺɀ Postulating the occurrence of certain accidents and verifying that, thanks to the planned functioning of the equipment, the consequences of these accidents will remain limited, is known as a deterministic approach. This approach is simple to apply in principle and allows an installation to be designed (and its ņŌņŇĸŀņ Ňł ĵĸ ņļōĸķҏ ŊļŇĻ ĺłłķ ņĴĹĸŇŌ ŀĴŅĺļŁņё ĵŌ ňņļŁĺ ņłіĶĴĿĿĸķ “envelope” cases. The deterministic approach is however unable to identify the most probable scenarios because it focuses attention on accidents studied with pessimistic hypotheses. The deterministic approach therefore needs to be supplemented by an approach that better reflects possible accident scenarios in terms of their probability, that is to say the probabilistic approach used in the “Probabilistic Safety Assessments” (PSA). Ļňņ ĹłŅ ŁňĶĿĸĴŅ ŃłŊĸŅ ŃĿĴŁŇņё ŇĻĸ ĿĸʼnĸĿ ӄׇ ĶłŁņļņŇ ļŁ establishing event trees for each “initiating event” leading to ŇĻĸ ĴĶŇļʼnĴŇļłŁ łĹ Ĵ ņĴĹĸĺňĴŅķ ņŌņŇĸŀ ҎĿĸʼnĸĿ ӆׇłĹ ĸĹĸŁĶĸ ļŁ ĸŃŇĻҏё defined by the failure (or the success) of the actions provided for in the reactor management procedures and the failure (or correct operation) of the reactor. The probability of each sequence is then calculated based on statistics on the reliability of systems and on the rate of success of actions (including data on “human reliability”). Similar sequences of events that correspond to the same initiating event are grouped into families, making it possible to determine the contribution of each family to the probability of reactor core meltdown. Although the PSAs are limited by uncertainties concerning the reliability data and approximations in the modelling of the facility, they consider a broader set of accidents than the deterministic assessments and enable the design resulting from the deterministic approach to be verified and supplemented if necessary. They are therefore to be used as a complement to deterministic studies and not as a substitute for them. The deterministic studies and probabilistic assessments constitute an essential element in the demonstration of nuclear safety that ĴķķŅĸņņĸņ ĸńňļŃŀĸŁŇ ļŁŇĸŅŁĴĿ ĹĴňĿŇņё ļŁŇĸŅŁĴĿ ĴŁķ ĸŋŇĸŅŁĴĿ ĻĴōĴŅķņё and plausible combinations of these events. To be more precise, the internal faults correspond to malfunctions, failures or damage to facility equipment, including as a result łĹ ļŁĴŃŃŅłŃŅļĴŇĸ ĻňŀĴŁ ĴĶŇļłŁє ŁŇĸŅŁĴĿ łŅ ĸŋŇĸŅŁĴĿ ĻĴōĴŅķņ correspond to events originating inside or outside the facility respectively and which can call into question the safety of the facility. Internal faults for example include: Ҋ loss of the electrical power supplies or the cooling systems; Ҋ ejection of a rod cluster control assembly; Ҋ rupture of a pipe in the primary or secondary system of a nuclear reactor; Ҋ reactor emergency shutdown failure. ļŇĻ ŅĸĺĴŅķ Ňł ļŁŇĸŅŁĴĿ ĻĴōĴŅķņё ŇĻĸ ĹłĿĿłŊļŁĺ ļŁ ŃĴŅŇļĶňĿĴŅ ŀňņŇ be considered: Ҋ flying projectiles, notably those resulting from the failure of rotating equipment; Ҋ pressure equipment failures; Ҋ collisions and falling loads; Ҋ explosions; Ҋ fires; Ҋ ĻĴōĴŅķłňņ ņňĵņŇĴŁĶĸ ĸŀļņņļłŁņђ Ҋ floods originating within the perimeter of the facility; Ҋ electromagnetic interference; Ҋ malicious acts. ļŁĴĿĿŌё ĸŋŇĸŅŁĴĿ ĻĴōĴŅķņ ŀłŅĸ ņŃĸĶļĹļĶĴĿĿŌ ĶłŀŃŅļņĸѓ Ҋ the risks induced by industrial activities and communication ŅłňŇĸņё ļŁĶĿňķļŁĺ ĸŋŃĿłņļłŁņё ĻĴōĴŅķłňņ ņňĵņŇĴŁĶĸ ĸŀļņņļłŁņ and airplane crashes; Ҋ earthquakes; Ҋ lightning and electromagnetic interference; Ҋ extreme meteorological or climatic conditions; Ҋ fires; Ҋ floods originating outside the perimeter of the facility; Ҋ malicious acts. Limiting the consequences of discharges On-site emergency plan Limiting the consequences of a severe accident Serious accident management Control of accidents Maintaining within the authorised range Design Operation Prevention of anomalies Regulation systems, periodic checks Backup systems, accident procedures Áǝƺ ژד ǼƺɮƺǼɀ Ȓǔ (ٹ ƺǔƺȇƬƺ ǣȇ (ƺȵɎǝ ٺ ASN Report on the state of nuclear safety and radiation protection in France in 2019 119 ׎ ٲ א THE PRINCIPLES OF NUCLEAR SAFETY AND RADIATION PROTECTION AND THE REGULATION AND OVERSIGHT STAKEHOLDERS 02