The need for strengthening of international cooperation in the area of analysis of radiological consequences

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1 ÚJV Řež, a. s. The need for strengthening of international cooperation in the area of analysis of radiological consequences Jozef Misak IAEA Technical Meeting on Source Term Evaluation of Severe Accidents October 2013, Vienna

2 Content The presentation summarizes the reasons for harmonization of acceptance criteria and methodology for assessment of radiological consequences of reactor accidents for various applications and provides relevant recommendations for the IAEA actions 1

3 Reference documents Safety Assessment for Facilities and Activities, GSR Part 4, IAEA (2009) Safety of NPPs: Design, SSR-2/1, IAEA (2012 Deterministic Safety Analysis for NPPs, SSG-2, IAEA (2009) Safety Assessment and Verification for NPPs, NS-G-1.2, IAEA (2001) Format and Content of SAR, GS-G-4.1, IAEA (2004) Reactor Harmonization Group, WENRA Reactor Safety Reference Levels, January 2008 WENRA, Reactor Harmonization Group, Reactor Safety Reference Levels, January 2008 European Utility Requirements for LWR NPPs. Rev. C, April

4 Applicability of radiological analysis Radiological analysis provides inputs for various documents developed and submitted for regulatory review in different stages of NPP life time, including o Different stages of Safety Analysis Reports o Environmental Impact Assessment o Emergency Preparedness and Response Programme o Environmental Monitoring Programme 3

5 Importance of harmonization Radiological consequences Represent the direct measure of the level of safety Are publicly sensitive issues and therefore influencing public trust Have trans-boundary effects and implications Are cross-cutting elements contained in several documents of the safety case Are important for international comparison of different reactor designs International harmonization of approaches to determination of radiological consequences is needed 4

6 Indication of areas where harmonization would be appropriate Differences in radiological acceptance criteria for design basis accidents Absence of radiological acceptance criteria for severe accidents Differences in methodology for demonstration of compliance with the criteria Internal inconsistencies in IAEA Safety Standards Differences between methodologies in IAEA Safety Standards and other documents (such as WENRA Reference Levels or European Utility Requirements) Differences in methodologies used in various licensing documents (EIA, SAR) 5

7 Large differences in radiological acceptance criteria for design basis accidents 6

8 IAEA SSR-2/1 on high level criteria Requirement 5: Radiation protection: The design of a nuclear power plant shall be such as to ensure that radiation doses to workers at the plant and to members of the public do not exceed the dose limits, that they are kept as low as reasonably achievable in operational states for the entire lifetime of the plant, and that they remain below acceptable limits and as low as reasonably achievable in, and following, accident conditions The design shall be such that for design basis accident conditions, key plant parameters do not exceed the specified design limits. A primary objective shall be to manage all design basis accidents so that they have no or only minor radiological impacts, on or off the site, and do not necessitate any offsite intervention measures.

9 Examples of radiological acceptance criteria for DBAs USA, Spain, Sweden, Korea, Japan,... Accident LOCA SGTR Main steam line break Locked rotor accident Rod ejection accident Fuel handling accident Small LOCA Gas waste system failure DEC or severe accidents Effective dose limit (at exclusion area boundary) 250 msv msv depending on additional conditions msv depending on additional conditions 25 msv 63 msv 63 msv msv depending on additional conditions 1 msv No limit established

10 Examples of radiological acceptance criteria for DBAs Germany, Slovakia, UK, Switzerland, Netherlands,... Germany. Slovakia 50 msv effective dose for all DBA UK, Switzerland, Netherlands (in some cases depending on frequency): 100 msv for frequency less than 1E-4/r.y In addition to different numbers attention should be paid to the fact that limits are prescribed for: Different duration of exposure Different pathways of exposure Different levels of conservatism in dose estimate => In many cases the criteria are too different and too large (not in compliance with IAEA Safety Standards), the first intervention level (sheltering) being ~ 10 msv in 2-7 days

11 Acceptance criteria for radioactive releases / max doses to general public (STUK, Finland) DBC 1, Normal operation radiation dose limit 0,1 msv / year for the entire site DBC 2, Anticipated events (f>1.e-2) radiation dose limit 0,1 msv DBC 3, Class 1 postulated accidents (1E-3 < f < 1E-2) radiation dose limit 1 msv DBC 4, Class 2 postulated accidents (f<1e-3) radiation dose limit 5 msv DEC, Design extension conditions, without core melt radiation dose limit 20 msv

12 Absence of radiological acceptance criteria for severe accidents 11

13 IAEA SSR-2/1 on high level criteria Requirement 5: Radiation protection: The design of a nuclear power plant shall be such as to ensure that radiation doses to workers at the plant and to members of the public do not exceed the dose limits, that they are kept as low as reasonably achievable in operational states for the entire lifetime of the plant, and that they remain below acceptable limits and as low as reasonably achievable in, and following, accident conditions The design shall be such that design extension conditions that could lead to significant radioactive releases are practically eliminated (see footnote 1); if not, for design extension conditions that cannot be practically eliminated, only protective measures that are of limited scope in terms of area and time shall be necessary for the protection of the public, and sufficient time shall be available to implement these measures.

14 Examples of acceptance criteria for severe accidents No quantitative radiological acceptance criteria established in majority of countries (Czech Republic, Slovakia, France, Germany, USA, Russia, etc) Requirements considered fulfilled if release is not more than 0,1 % of the core inventory of the caesium isotopes 134 and 137, contained in a reactor core of 1800 MWth (Sweden) Maximum release of Cs TBq (Finland) Atmospheric release of caesium-137 below 30 TBq and the combined fall-out of nuclides other than caesium-isotopes shall not cause, in the long term, starting three months from the accident, a hazard greater than would arise from a caesium release corresponding to the above-mentioned limit (EUR, Bulgaria) EUR targets for short term and long term actions and for limited economic impact 13

15 EUR Targets for short term protective actions Duration Distance Target Objective and observation Target for emergency actions 7 days from the release initiation from the plant Any distances Effective dose committed 50 msv It s intended to assure that at any distances from the plant, emergency evacuation of the public is not required. Cumulated releases during first 24 hours of accident are considered. Exposure by irradiation from the plume, from deposits and from inhalation should be considered (not from ingestion). Target for delayed actions First 30 consecutive days after the release termination Beyond 3 km Effective dose committed 30 msv It is intended to assure that beyond 3 km from reactor public evacuation within 30 days after termination of release is not required. Cumulated releases during first 4 days of accident are considered. Exposure by irradiation from the deposits and from inhalation due to resuspension of deposits should be considered (not from ingestion). Targets for protective actions are individual dose limits

16 EUR Target for long term protective actions Duration Distance Target Objective and observation Target for long-term actions Up to 50 years from the termination of all releases Any distances Effective dose committed 100 msv It is intended to assure that at any distances from the plant public relocation after the release termination is never required. Exposure by irradiation from the deposits and from inhalation due to resuspension of deposits should be considered (not from ingestion). Targets for protective actions are individual dose limits

17 EUR Targets for economic impact Duration Distance Target Objective and observation 1 st Target for Economic Impact 2 nd Target for Economic Impact After 1 year from the end of the accident After 1 month from the end of the accident Beyond 10 km Beyond 100 km 1250 Bq/kg for Cs Bq/kg for I Bq/kg for Sr 90 This land contamination limit would allow free trading of crops cultivated beyond the said distance from the reactor according to existing EC regulations. The limit is based on a dose of 5 msv to individuals eating contaminated food for 1 year. Targets for economic impacts are land contamination limits

18 Differences in methodology for demonstration of compliance with criteria among the countries 17

19 Differences in methodology of analysis of consequences among the countries Large difference in determination of core inventory fractions released under DBA conditions between the US (and Japan, Korea, or Spain) and majority of European countries USA (RG 1.183): the release of iodine and noble gases starts with gap inventory, continuing with releases from the fuel matrix, assuming that the core will melt and releases from the molten corium will take place even in the case of DBA In many European countries it is assumed that only a fraction of fuel will fail releasing gap inventory to the RCS Conservative approach to prediction of the limited number of the fuel elements failed is used in accordance with the regulatory guidance documents in Finland, UK, France, Russia, Slovakia, etc. 18

20 Differences in methodology of analysis of consequences among the countries EPR methodology: The assessment of released activity is based on conservative methods and assumptions (initial primary activity, rate of cladding failures, etc). The assumptions for calculating the radiological consequences (evaluation of doses) are set realistically in order to arrive at a reasonably conservative assessment of the radiological consequences Other methodological assumptions: the level of reactor coolant activity and the treatment of iodine spiking in DBAs, the inventory of fission products in the gap, forms of iodine and others 19

21 Inconsistencies in IAEA Safety Standards 20

22 Safety Requirements and Safety Guides for design, for safety assessment, for content of SAR GSR Part 4, art The aim of the deterministic approach (in safety analysis) is to specify and apply a set of conservative deterministic rules and requirements This conservative approach provides a way of compensating for uncertainties GSR Part 4, Requirement 17: Uncertainty and sensitivity analysis shall be performed and taken into account in the results of the safety analysis and the conclusions drawn from it. SSR-2/1, art The design basis accidents shall be analysed in a conservative manner. SSR-2/1, art. 5.27: The effectiveness of provisions to ensure the functionality of the containment (in case of design extension conditions) could be analysed on the basis of the best estimate approach NS-G-1.2, art. 4.19: In general, the deterministic analysis for design purposes should be conservative. The analysis of beyond design basis accidents is generally less conservative than that of design basis accidents. 21

23 Safety Requirements and Safety Guides for design, for safety assessment, for content of SAR GS-G-4.1, art Deterministic analysis It is acceptable that best estimate codes are used for deterministic analyses provided that they are either combined with a reasonably conservative selection of input data or associated with the evaluation of the uncertainties of the results. GS-G-4.1, art The analyses (of beyond design basis accidents) should use best estimate models and assumptions and may take credit for realistic system action and performance, SSG-2, art Although conservative assumptions and bounding analyses should be used for design purposes, more realistic analyses should be used to evaluate the evolution and consequences of accidents SSG-2, chapter 9 specifically dealing with source term evaluation does not provide any guidance on the use of conservative vs realistic analysis => Very limited and not always clear guidance on radiological analysis is provided in the IAEA Safety Standards 22

24 Differences between methodologies in IAEA Safety Standards and other documents (such as WENRA Reference Levels or European Utility Requirements) 23

25 Conservative or best estimate analysis in various documents GSR Part 4, art The aim of the deterministic approach (in safety analysis) is to specify and apply a set of conservative deterministic rules and requirements SSR-2/1, art The design basis accidents shall be analysed in a conservative manner. WENRA Reference Levels, E 8.1: The initial and boundary conditions (in safety demonstration for design basis accidents) shall be specified with conservatism. WENRA Reference Levels, F 2.2: Realistic assumptions and modified acceptance criteria may be used for the analysis of the beyond design basis events. 24

26 Conservative or best estimate analysis in various documents EUR, section Deterministic analysis, item 14)... For calculation of releases, physically-based assumptions and best estimate evaluations, with suitable margins to take into account uncertainties, are preferred. => there is not full consistency among various documents 25

27 Differences in methodologies used in various licensing documents (EIA, SAR) 26

28 Inconsistency between radiological analysis in different licensing documents In several countries, different licensing documents reviewed by the same regulatory body are developed using different approaches to radiological analysis This is in particular true for Safety Analysis Reports and Radiological Environmental Impact Reports Even the analysis of the same accidents (DBAs, severe accidents) uses very different assumptions (e.g. scope of the core damage, weather conditions) Subsequently, the results of analysis in terms of doses may differ by one or two orders Since both documents are publicly available, such different publicly sensitive information may be confusing for the public 27

29 Conclusions Radiological assessment is a key component for overall assessment of safety of NPPs There are significant differences among the countries in acceptance criteria and methodology for assessment of radiological consequences, in particular in reactor accidents Acceptance criteria for DBAs are often not in compliance with IAEA standards, criteria for severe accidents rarely defined The issue of radiological consequences is not sufficiently covered in existing IAEA Safety Standards and other guidance documents Harmonization of criteria and methodology for radiological assessment can contribute to consistency of information about safety of NPPs It is understood that harmonization may be a difficult process due to the close interrelation of the issue with national legislation, but appropriate IAEA Safety Standards can become a driving force towards better harmonization 28

30 Recommendations for the IAEA More attention should be devoted by the IAEA to the issue of assessment of radiological consequences of reactor accidents In ongoing revisions of the IAEA Safety Standards attention should be paid to ensuring consistency and comprehensiveness in addressing radiological consequences in various standards Publication of relevant IAEA lower level technical documents should be accelerated Development of a specific Safety Guide (or updating of existing ones) on assessment of radiological consequences should be considered 29