Ř the need for strengthening of international cooperation
TRANSCRIPT
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
21 – 23 October 2013, Vienna
1
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
2
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 2001
3
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
4
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
5
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)
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.
5.25. 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 off-site intervention measures.
Examples of radiological acceptance criteria for
DBAs – USA, Spain, Sweden, Korea, Japan, ...
AccidentEffective dose limit (at exclusion area
boundary)
LOCA 250 mSv
SGTR 25-250 mSv depending on additional conditions
Main steam line break25-250 mSv depending on additional conditions
Locked rotor accident 25 mSv
Rod ejection accident 63 mSv
Fuel handling accident 63 mSv
Small LOCA25-250 mSv depending on additional conditions
Gas waste system failure 1 mSv
DEC or severe accidents No limit established
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 exposureDifferent pathways of exposureDifferent 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
Examples of radiological acceptance criteria for
DBAs – Germany, Slovakia, UK, Switzerland,
Netherlands, ...
Acceptance criteria for radioactive releases / max doses to general public (STUK, Finland)
DBC 1, Normal operationradiation 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 meltradiation dose limit 20 mSv
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.
5.31. 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.
13
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 137 100 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
EUR Targets for short term protective
actions
Duration Distance Target Objective and observation
Ta
rget
fo
r
emer
gen
cy a
ctio
ns
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).
Ta
rget
fo
r
del
ay
ed a
ctio
ns
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
EUR Target for long term protective
actions
Duration Distance Target Objective and observation
Ta
rget
fo
r lo
ng
-ter
m
act
ion
s 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
EUR Targets for economic impact
Duration Distance Target Objective and observation
1st
Ta
rget
fo
r
Eco
no
mic
Im
pa
ct
After 1 year
from the end of
the accident
Beyond 10 km 1250 Bq/kg for
Cs137
2000 Bq/kg for
I131
750 Bq/kg for
Sr90
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.
2n
dT
arg
et f
or
Eco
no
mic
Im
pa
ct
After 1 month
from the end of
the accident
Beyond 100 km
Targets for economic impacts are land contamination limits
18
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.
19
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
21
Safety Requirements and Safety Guides for design,
for safety assessment, for content of SAR
• GSR Part 4, art. 4.54.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. 5.26. 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.
22
• GS-G-4.1, art. 3.128. 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. 3.140. 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. 3.8. 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
Safety Requirements and Safety Guides for design,
for safety assessment, for content of SAR
23
Differences between
methodologies in IAEA Safety
Standards and other documents
(such as WENRA Reference
Levels or European Utility
Requirements)
24
Conservative or best estimate analysis in
various documents
• GSR Part 4, art. 4.54.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. 5.26. 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.
25
Conservative or best estimate analysis in
various documents
• EUR, section 2.1.3.3 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
27
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
28
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
29
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