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.

2

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

3

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

4

・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.

5

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

6

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

7

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.

8

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

10

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)

11

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

12

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.

13

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

14

Full-scope comprehensive site Level 3

NRC ordered a full-scope, site comprehensive Level 3 PRA

（SECY-11-0089）

Vogtle Unit 1&2

15

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

16

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.

17

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.

18