LES ACTIVITÉS NUCLÉAIRES : RAYONNEMENTS IONISANTS ET RISQUES POUR LA SANTÉ ET L’ENVIRONNEMENT The final result is a prudent over-estimation, which gives an approximate idea of the doses liable to be received per category of sanitation workers, according to the facility which carries out the discharge, the collection system receiving it and the plant which treats the wastewater. CIDRRE is a means of ensuring that the annual dose received by the sanitation workers remains below 1 mSv. For the population, the estimated radiological impact linked to radioactive discharges from nuclear medicine units and research laboratories into the sewerage systems would appear to be below 300 μSv/year in all the studies, even in worst-case scenarios considering all the radionuclides detected in the sanitation systems. This impact is estimated to be lower than 1 μSv/year when the radionuclides used in nuclear medicine are not taken into account, with realistic hypotheses (IRSN data). The radiological impact of BNIs In accordance with the optimisation principle, the licensee must reduce the radiological impact of its facility to values that are as low as possible under economically acceptable conditions. The licensee is required to assess the dosimetric impact of its activity. As applicable, this obligation is the result of Article L. 1333‑8 of the Public Health Code, or the regulations concerning BNI discharges (Article 5.3.2 of amended ASN resolution 2013-DC-0360 of 16 July 2013 concerning control of detrimental effects and the impact of BNIs on health and the environment). The result must be compared with the annual dose limit for the public (1 mSv/year) defined in Article R. 1333‑11 of the Public Health Code, which corresponds to the sum of effective doses received by the public as a result of nuclear activities. In practice, only traces of artificial radioactivity are detectable in the vicinity of the nuclear facilities; most measurements taken during routine surveillance are below the decision threshold or reflect the natural radioactivity. As these measurements cannot be used for dose estimations, models for the transfer of radioactivity to humans must be used, on the basis of measurements of discharges HIGHLIGHT No. 7 Inspections with sampling and measurements Inspections involving sampling and measurement, most of which are carried out unannounced, contribute to ASNR’s monitoring activities to ensure that licensees comply with the regulatory provisions they are required to comply with in terms of controlling discharges and monitoring the environment at their facilities. In particular, the representativeness of the samples, the quality of the analyses and the analytical performance of the licensee’s laboratory were questioned on this occasion, through contradictory measurements. Since 1 January 2025, all sampling and analysis of radioactivity in effluent and the environment has been carried out by ASNR. This type of inspection is carried out at all nuclear installations, according to a multi-year schedule tailored to the issues at stake at each facility. ASNR carries out around 15 sampling inspections a year, with the aim of ensuring that each NPP site is inspected and sampled at least once every two years, or even once a year for the “fuel cycle” facilities with the highest stakes. These inspections also contribute to acquiring knowledge of the characteristics of discharges, monitoring the radiological state of the environment and maintaining the skills of ASNR’s experts in taking samples and carrying out analyses – essential skills that would be called upon heavily in the event of an emergency. The inspection team generally consists of two inspectors and two experts in radiological sampling and analysis. The pool of these experts is now made up of around thirty people to meet the new organisational requirements following the creation of ASNR. The organisation of an inspection with sampling and analysis follows the process of an inspection, with the following specific stages: • the preparation of a sampling plan in advance of the inspection, identifying the sampling locations and the analyses to be carried out on each sample; • the collection of samples from liquid or gaseous effluent discharges on site and in various environmental compartments (atmospheric, aquatic or terrestrial) during normal operation of the facilities; • the analyses are then carried out independently by the licensee and the ASNR laboratories, with both parties involved. These analyses concern the determination of natural and artificial radioactivity and generally relate to the activity of tritium, carbon-14, gamma emitters, or the determination of total alpha or beta radioactivity indices. More specific determinations such as the activity of nickel-63, strontium-90 and uranium isotopes can also be carried out; • a follow-up letter, including in particular the analyses to be carried out along with deadlines, requests for corrective action and additional information with regard to the deviations observed during the inspection, is sent to the licensee and posted on the ASNR website; • the results of the radiological analyses obtained by the licensee are compared with the regulatory requirements (discharge limits, performance of the analysis technique in terms of decision threshold values and uncertainties, etc.) and with the results obtained by ASNR according to defined criteria. Deviations are the subject of a second follow-up letter if necessary. In 2025, ASNR carried out 14 inspections involving sampling and measurement, with around 70 sampling points resulting in more than 300 radiological analyses, i.e. two-thirds of the different types of analysis (physico-chemical, PFAS, bacterial, etc.). This inspection system supplements the results of environmental monitoring recorded in the monthly registers that licensees send to ASNR, the results of contradiction analyses carried out as part of “cross-checks” and inspections by laboratories approved by ASNR as part of the national Network for Measuring Radioactivity in the environment (RNM). HIGHLIGHT No. 8 With regard to measurements The Decision Threshold (SD) is the value above which it is possible with a high degree of confidence to conclude that a radionuclide is present in the sample. The Detection Limit (LD) is the value as of which the measurement technique is able to quantify a radionuclide with a reasonable degree of uncertainty (the uncertainty is about 50% at the LD). More simply, LD ≈ 2 x SD. For the measurement results on chemical substances, the Quantification Limit is equivalent to the Detection Limit used to measure radioactivity. REFERENCE SPECTRA For the NPPs, the reference spectra of discharges comprise the following radionuclides: • Liquid discharges: tritium, carbon-14, iodine-131, other fission and activation products (manganese-54, cobalt-58, cobalt-60, nickel-63, silver-110m, tellurium-123m, antimony-124, antimony-125, caesium-134, caesium-137); • Gaseous discharges: tritium, carbon-14, iodines (iodine-131, iodine-133), other fission and activation products (cobalt-58, cobalt-60, caesium-134, caesium-137), noble gases: xenon-133 (continuous discharges from ventilation networks, draining of ‘RS’ effluent storage tanks and during the depressurisation of reactor buildings), xenon-135 (continuous releases from ventilation systems and during the decompression of reactor buildings), xenon-131m (draining of RS tanks), krypton-85 (draining of RS tanks), argon-41 (during the decompression of reactor buildings). ASNR Report on the state of nuclear safety and radiation protection in France in 2025 163 01 02 03 05 09 06 10 07 11 13 04 08 12 A / Z
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