The Current State of Level 3 PSA...The Current State of Level 3 PSA Toshimitsu Homma Nuclear Safety...
Transcript of The Current State of Level 3 PSA...The Current State of Level 3 PSA Toshimitsu Homma Nuclear Safety...
The Current State of Level 3 PSA
Toshimitsu Homma
Nuclear Safety Research Center
Japan Atomic Energy Agency
Document 3 , The 3rd Meeting,
Working Group on Voluntary Efforts and Continuous
Improvement of Nuclear Safety,
Advisory Committee for Natural Resources and Energy
The Outcome of the IAEA Technical Meeting (July 2012)
The current state of implementation of Level 3 PSA (L3PSA) The implementation of L3PSA is limited as compared with L1 & 2 PSA, and
moreover, most cases of its implementation are 10 to 20 years old. A few recent
examples include Koeberg NPP (South Africa) L3PSA and the L3PSA study project
(United states) which is now underway.
The current state of methodology The general method has been adequately established. The IAEA document (1992)
should be improved by adopting the latest technology and the lessons learned from
the accident in Fukushima. Such improvements will affect the model and the
approach associated with L3PSA. In addition, some calculation codes may already
have become obsolete.
Research and development The project currently under way will provide the latest standard method (ASME PRA
Standard) and will serve as an example of the capacity of L3PSA to deal with the
current issues. Research is in progress in other technological fields (nuclide
migration in the environment and economic impacts).
Direction of L3PSA L3PSA is not practiced in many countries because no strict regulatory requirements
exist. Nevertheless, the participants identified many useful applicable fields for
regulation.
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The Background of L3PSA
Various L3PSA codes have been developed in Europe and in the United States since the WASH-
1400 Report of the United States in 1975, and they were used for the risk assessment of severe
accidents.
The L3PSA section (probabilistic accident consequence analysis ) exists as an extension of both the
conventional exposure dose assessment, and the assessment of radioactive nuclide migration and
dose in the environment, which constitutes the major portion of the foregoing, and is a well
established technology supported by research on nuclear test fallout, the Chernobyl accident, etc.
The scope of application is not confined to risk assessment but has been expanding to areas
including the development of emergency plans and the assessment of nuclear externality (external
costs).
1970 1980 1990 2000
第1回比較計算 第2回比較計算 専門家判断P
WASH-1400 NUREG-1150米国 CRAC CRAC2 MACCS MACCS2
Sizewell PWR Hinkley Pointイギリス MARC MARC2A
CONDORGerman Risk Study
ドイツ UFOMOD New UFOMOD
EC COSYMAPC-COSYMA
モデルプラントPSA原研 OSCAAR
US
UK
Germany
JAERI
1st comparison calculation 2nd comparison calculation; expert judgment P
Model plant PSA
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Procedure of L3PSA (Probabilistic Consequence Analysis)
Atmospheric
dispersion analysis
Analysis of
ground depositions
Evaluation of
early exposure
doses
Evaluation of
long-term
exposure doses
Sampling of
meteorological
data
Health effects
assessment
Economic loss
evaluation
Analysis of the
effect of measures
aimed at reducing
exposure dose
Meteorological
data
Emission
source
information
Distribution
of the population,
and of agricultural
and livestock
products
Composition of OSCAAR Codes
iA )(, rji
Concentrations
・Atmospheric
・Ground-level
Doses
・Organ and tissue doses
・Effective doses
Effects
・Number of early deaths
・Number of cancer deaths
)(, rD ji
)(, rC ji
NiCpsR iii ,...,2,1,,,
Level 2 PSA
・Accident scenario:
・Event probability:
・Source term:
is
ip
iA
About 60
different nuclides
Various meteorological
conditions All exposure pathways
Realistic
assessment
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・Cloud shine ・Ground shine ・Inhalation/intake(food) ・Inhalation of re-airborne materials
International Comparative Calculation of L3PSA Codes Outline
Implemented from 1991 through 1994 and co-organized by the OECD/NEA and the CEC.
The main objective was to compare the predicted results among the participated codes and
clarify the causes of differences, and at the same time, to use it for QA (quality assurance) of
the codes.
The object of comparison is the assessment procedures under multiple calculation conditions
including five source terms and existence/nonexistence of protective measures targeted at a
virtual site in Europe. The results are compared in a probabilistic form (CCDF).
Participated codes
ARANO (VTT, Finland), CONDOR (SRD and NRPB, UK), COSYMA (KFK, Germany, and
NRPB, UK), LENA (SSI, Sweden), MACCS (SNL, US), and OSCAAR (JAEA, Japan)
Conclusion and recommendations
The degree of difference in the evaluation results among the codes varied depending on the
evaluation item, but the difference was generally within several factors. The main cause of
differences is attributed to the model structure and the hypothesis employed.
The difference in evaluation results among the codes is small when compared with the total
uncertainty associated with risk estimation, and it will not bring about obstacles in using any
of the participated codes in a comprehensive risk assessment
For the future, it is necessary to understand the uncertainty associated with the result of a
Level 3 PSA Code and its relative contribution to the uncertainty associated with the overall
risk estimation.
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Result of Comparison among Codes (example)
Distance from the point of release (km)
条件
付早
期死
亡確
率
Collective effective dose commitment (man Sv)
条件
付発
生確
率
No protective measures
(Example)
Calculation task 1
・Source term
Time before release : 2 hr.
Duration of release: 1 hr.
Release height:10m
Warning time:1 hr.
Release fraction: Xe-Kr:
1.0, Org-I: 0.001, I:0.1,
Cs-Rb: 0.1, Te-Sb: 0.1,
Ba-Sr, Ru: 0.01, La: 0.001
・Protective measures
Sheltering, evacuation,
relocation, food restriction
Con
ditio
nal p
rob
abili
ty o
f o
ccu
rre
nce
Co
nd
itio
nal p
rob
abili
ty o
f e
arly d
ea
ths
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Target for evaluation Name of project Outline Implementing and participating
agencies
Validation of a
migration model
BIOMOVS
(1986-1990,1992-1996)
Validation of the model for transfer of radionuclides in the
biosphere on the basis of actual measured data.
・Contamination of crop plant with131I and 137Cs due to the
Chernobyl accident
SSI, Sweden; JAERI, Japan;
ORNL, US; NRPB, UK; GSF,
Germany; etc.
Validation of a
migration and dose
evaluation model
IAEA-VAMP
(1988-1994)
Validation of the model for transfer of radionuclides in terrestrial
parts of the earth, an urban region and the hydrosphere
・A transfer path and dose assessment for 137Cs caused by the
Chernobyl accident (in Central Bohemian Region and Finland)
IAEA;
CRNL, Canada; PNL, US;
MAFF, UK; GSF, Germany; etc.
Code comparison NEA/CEC international
comparison calculation
(1991-1994)
Comprehensive code comparison of Level 3 PSA codes JAERI, Japan; VTT, Finland;
SSI, Sweden; SRD and NRPB,
UK; KFK, Germany; SNL, US
Uncertainty evaluation EC/NRC uncertainty
evaluation (1993-1999)
A collection of information on expert judgment about the
parameters related to the environmental impacts assessment and
the evaluation of uncertainty associated with the input parameters
for Level 3 PSA
EC; NRC, US
Validation of a
migration and dose
evaluation model
IAEA-BIOMASS
(1996-2001)
Dose reconstruction relating to environmental releases in the past
・Accidental release of 131I from Hanford Reprocessing Plant
・Contamination of the Iput river basin with137Cs due to the
Chernobyl accident
IAEA
JAERI, Japan; PNL, US;
MAFF, UK; IFE, Norway; etc.
Validation of a
migration and dose
evaluation model
IAEA-EMRAS
(2002- )
Release of 131I due to the Chernobyl accident
・Thyroid burden in the Tula region, Russia
・The effect of distribution of stable iodine tablets in Poland
IAEA
JAERI; BNFL, UK; NCI, US;
CLRP, Poland; etc.
Reliability Evaluation Study relating to L3PSA
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Uncertainty Evaluation concerning L3PSA Project concerning EC/USNRC’s expert judgment information (1993-1999)
Background: The severe accident risk assessment in the U.S. (NUREG-1150) was intended to
evaluate uncertainty comprehensively, but did not include any uncertainty evaluation of Level 3 PSA.
About 70 experts from Europe and the United States participated in 8 expert panels (atmospheric
diffusion, deposition, external exposure, internal exposure, food chain (direct) , food chain (indirect),
early effects, and late effects ).
While the uncertainty associated with the input parameters of COSYMA and MACCS codes was the
object of the evaluation, the distribution of the uncertainty related to each parameter defined by
asking questions to experts about the distribution of uncertainty of the quantities which are not
dependent on the model and which are fundamentally measurable (e.g. , the uncertainty related to
metabolic parameters is obtained from the uncertainty related to nuclide residual volume in the
body).
Distance from the site (km)R
isk
of e
arly d
eath
Distance from point of release (km)
Con
ditio
nal p
roba
bilit
y of
av
erag
e in
divi
dual
ear
ly d
eath
s
Uncertainty evaluation on individual risks
(A calculation example by JAERI)
The uncertainty evaluation for EC COSYMA code
Uncertainty analysis is performed for each individual
process of atmospheric diffusion/deposition, for dose
calculation, food chain, and health effects on the principal
parameters.
The uncertainty analysis of Level 3 PSA as a whole is
conducted and uncertainties are quantified with respect
to individual evaluation items.
The parameters that contribute to the uncertainty of the
evaluation result are identified.
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Establishment of safety goals
Optimization of the safety margins of regulations
Establishment of effective
disaster prevention plans
Regulations based on a concept
common to various facilities Nuclear reactor facility
Fuel cycle facility
Waste treatment and disposal
More reasonable regulations
Ris
k info
rmation Validity assessment of the measures
against severe accidents
Study of the measures reducing the
risks associated with earthquakes
Study of rational safety management
Rational design of the next-generation
of reactors
Further safety enhancements
Study of the nuclear damage
compensation system
Evaluation of the externality of
energy sources
Comparative risk evaluation
Application field
Technological study based on probabilistic safety assessments (PSA)
Utilization of Risk Information 9
Development of Safety/Performance Objectives
0.01
0.1
1
10
100
1000
0.01 0.1 1 10
セシウム類の炉内蓄積量に対する放出割合(%)
137C
sが14
80kB
q/m
2を超
える
領域
面積
(km
2)
95%値平均値
50%値
1.0E-04
1.0E-03
1.0E-02
1.0E-01
1.0E+00
0.1% 1.0% 10.0% 100.0%
揮発性物質の放出割合
条件
付平
均が
ん死
亡確
率 評価範囲 : 1km
評価範囲 : 2km
評価範囲 : 10km
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Quantitative objective: First and foremost, mitigation of risk should mean mitigation
of various types of risk according to the frequency of occurrence. Mitigating risk
against an individual will lessen the other types of risk at the same time though the
levels of mitigation may not be appropriate.
Health risk to a population: Frequency of accidents which result in considerable
damage should be mitigated according to the size of the damage. Also, this principle
should be applied to characteristics and locations of facilities.
Societal risk: It is hard to quantify the societal influence. Moreover, discussion on to
which level of risk we should aim to mitigate has made little progress.
Relationship between release and area of moving
(NSC safety goal expert committee, 17-3)
Conditional probability of death
(NSC performance goals expert committee, 7-3)
Proportion of Cs released vs Cs accumulated in the reactor (%) Proportion of volatile substances released into the atmosphere (%)
average value
value
Evaluation area
Evaluation area
Evaluation area
Distance from the point of release (km)
0.001
0.01
0.1
1
0.1 1 10 100
後期大規模放出
9×10-8
/炉年
実効線量 50mSv
早期大規模放出
2×10-10
/炉年
骨髄線量 1Sv
各線
量を
超え
る気
象の
出現
確率
Study on Nuclear Emergency Preparedness
More rational protective measures are to be examined on the basis of the findings from the
studies including the PSA study, the study on severe accidents, and the study of methods of
optimizing protective measures.
Study on emergency planning zone (EPZ)
Pro
ba
bili
ty o
f o
ccu
rre
nce
of clim
atic c
on
ditio
n
wh
ere
ind
ivid
ua
l d
oses a
re e
xce
ed
ed
Large late release
9x10 – 8 / reactor year
Effective dose: 50mSv
Large early release
2x10-10 / reactor year
Bone-marrow dose : 1Sv
Analysis of effects on dose reduction
(The 2nd Proactive emergency plan project team)
No action
Sheltering(for 2 days)
Sheltering in concrete buildings (2days) + evacuation(7days)
preventive evacuation
Distance from the point of release (km)
Th
yro
id e
qu
iva
len
t d
ose
(S
v)
Reduction with KI
(taken within 12 hours after release)
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Application of L3PSA in the United States
Reactor Safety Study (WASH-1400, 1975)
Technical Guidance for Siting Criteria Development (NUREG/CR-2239,
1982)
Severe Accident Risk (NUREG-1150)
Reassessment of selected factors affecting siting (NUREG/CR-6295, 1997)
Study of protective action recommendation (PAR) (NUREG/CR-6953, Vol.
1, 2007)
State-of-the-Art Reactor Consequence Analyses (SOARCA) (NUREG-
1935, 2013) (NUREG/CR-7110 Vol. 1 & 2, 2013)
SOARCA uncertainty analysis
Scoping analysis of spent fuel pool
Analysis of filtered containment vent
Comprehensive site level 3 PRA
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Application of the Guidance for Short-term Protective Actions (NUREG-0654, Rev.1, draft version, 2009)
In 2004, NRC made a study
on alternative methods to
replace conventional
protective actions (PAR
Study, NUREG/CR6953
Vol.1).
As a result, it became
necessary to revise
NUREG-0654, Rev.1,
Supp.3.
The evacuation time
estimate(ETE) that must
be assessed in the
development of an
emergency plan is
considered in the
implementation of
protective actions.
Large early
release?
Is 3 hr. or more of ETE required for 90% completion of the evacuation in areas
within 2-mile radius?
Do GE conditions
continue to exist?
Do conditions for a
serious emergency
continue to exist?
Obstacles to
evacuation? Are obstacles
removed?
Declaration of general
emergency (GE)
Sheltering on the spot: in areas within a 2-mile radius; evacuation in areas within2-5 miles leeward.; sheltering on the spot: in areas within 5-10 miles
To be implemented in case the
evacuation in areas within a 2-
mile radius can be safely done.
Continuation of
the assessment
Evacuation is implemented
in areas within 2-5 miles
leeward at T=X*
Evacuation in areas within 2-
mile radius; sheltering on the
spot in areas within 5 miles
leeward. Make preparations in
other areas.
Sheltering indoors on the spot in areas within a 2-mile radius; make preparations in areas within 5 miles leeward and in other areas.
Continue the evaluation
to maintain PAR
Confer with ORO on the
maintenance or the
expansion of PAR
Yes
Yes
Yes
Yes
Yes
Yes
No
No
No
Yes
No
No
No
*X means ETE for 90% completion of the evacuation
in areas within 2-mile radius.
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SOARCA (NUREG-1935, 2012)
Optimum evaluations of off-site radiation health effects caused by severe
accidents at 2 plants (Peach Bottom, Surry) to be conducted.
Objective
To revise evaluations of effects caused by severe accidents (especially, location research
in 1982)
To reflect the present status of plants and improvements made related to nuclear security
To utilize up-to-date models (MELCOR, MACCS2)
To improve communication with various stakeholders
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Full-scope comprehensive site Level 3
NRC ordered a full-scope, site comprehensive Level 3 PRA
(SECY-11-0089)
Vogtle Unit 1&2
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Objectives
To reflect technical progress made
in PRA modeling, plant operations,
safety and security. (multi-units,
accidents at SFP, accidents related
to dry casks)
To improve NRC staff’s ability to
conduct a PRA
To specify new information to
strengthen regulatory decision
making processes
To evaluate technical feasibility and
cost as to L3PRA
Application Field of L3PSA
Application of the regulations
Establishment of the complete scope of safety goals and how to
satisfy them
Risk communication with other national agencies (in charge of
the protection of the environment and emergency preparedness)
and with the public
Priority of safety issues
Application in the industry
Decision support concerning design changes and operation
Optimization of severe accident management strategies
Emergency plans and their execution
Preparation of environmental impact assessment reports
Establishment of financial indemnity limits
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Benefits of the Implementation of L3PSA
L3PSA quantifies the off-site radiological consequences of a nuclear accident,
and enables a comprehensive understanding of the risk associated with
nuclear facilities. Relatively small cost is required to undertake this EPA
compared with L2PSA.
L3PSA is not only very important for risk communication with other fields
(environmental protection and emergency response) but also useful for risk
communication with the public (as compared with risk indices in L1 & 2PSA).
L3PSA supports the optimization of accident management plans,
preparedness and response to emergency situations by utilizing risk
information.
L3PSA is capable of providing useful information concerning the
development of responses to actual emergency situations.
L3PSA is useful for the decision-making concerning the siting of nuclear
facilities through the utilization of relevant risk information.
Also, it can provide information useful for land-use plans and for infrastructure
development in the neighborhood of the site.
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Level 3 Risk Index
The Risk index used in Level 3 PSA address various consequences of accidents such as
health effects (individual/collective, short-term/long-term doses), environmental impacts
(media contamination), and the economic impact (area and the population for protective
action, control of agricultural and livestock products).
Risk index must be meaningful and useful to the regulator, industry, other government
agencies, international organizations and to the public. Choice depends on the object.
Important information relating to decision-making, including environmental protection,
preparedness and responses to emergency plans, and land utilization programs, can be
provided.
Risk communication using the risk index is likely to raise concerns about safety, to increase
the sense of responsibility of the industry, and to contribute to the improvement of the
safety culture.
Risk index is useful for risk communication with the public, however, attention should be
paid to how it should be used during discussions.
Non-radiation effects are significantly larger than radiation effects as with the cases of
Chernobyl and Fukushima.
Expected value or risk curve (spectrum of frequency/consequences)?
The low frequency / large consequence nature of the problem needs to be addressed.
The Uncertainty inherent in the Risk Index is an issue that needs to be addressed.
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