WP 5 Safety and Transient Analysis Task 5.6 Containment and source term assessment for the ETDR LEADER Lead-cooled European Advanced DEmonstration Reactor

WP 5 Safety and Transient AnalysisTask 5.6 Containment and source term assessment for the ETDR

LEADERLEADERLLead-cooled ead-cooled EEuropean uropean AAdvanced dvanced DEDEmonstration monstration RReactoreactor

WP 5 - MEETINGWP 5 - MEETING

26th February 201326th February 2013


26 Feb 2013, Petten26 Feb 2013, Petten 22

Participants Royal Institute of Technology – Stockholm (KTH) Empresarios Agrupados (EA)

Objective Provide scoping assessment of the source term available for

release to the environment Thermodynamic calculations Containment leakage and rates Calculations of fission products formed

Assess the transport of airborne radionuclides in containment

Perform calculations of the radiological impact of the releases to the public



Second Part: In charge of EA Perform the radiological analysis to determine the impact on

the environment Define the methodology to calculate the radiological impact Define the dose limits for each condition

Input needed from: The first part of this Task: Source Term Assessment by KTH WP 3 Conceptual Design, to define the release mode and the

meteorological conditions

Dose assessment for accident condition



Dose assessment methodology: A preliminary dose assessment will be performed with the

following objectives:

To calculate a preliminary site boundary dose to public (at the Exclusion Area Boundary) based on the reference source terms with no credit to any engineered safeguards.

To verify the compliance with the present European criteria for licensing LWR

To analyse the sensitivity on the dose to the public of the main uncertainties: retention of radionuclides in the primary coolant, fuel damage extension,...

To identify radionuclides with the biggest influence on the environmental dose and to highlight the systems able to limit their release.

To analyse the impact on the results of the dose calculation methodology used.



The calculation of accident conditions will be processed according to the Regulatory Guide 1.183 issued by the US NRC

In the calculation process, the ways of contamination

considered are:

The external exposure by submersion in a radioactive cloud.

The internal exposure by inhalation of radioactive particles.

No credit is given to the exposure via ingestion of contaminated food (milk, plants, meat…).



The dose limits for members of the public set in case of accident condition:

The effective dose limits for members of the public set in case of accidents are (NRC RG 1.183):

250 mSv for the accidents assumed to have the worst radiological consequences: a LOCA, a main steam line break for a PWR, a steam generator tube rupture for a PWR…, etc

63 mSv for accidents with less radiological impact: BWR Rod Drop Accident, PWR ROD Ejection Accident…, etc

The new trends for the dose release targets of currently designed power plants are around 50 mSv for accidents

Public

Whole body(External exposure)

250 mSv(R.G. 1.195)

Thyroid(Internal exposure)

3000 mSv(R.G. 1.195)

TEDE (CEDE+DDE)250 mSv

(R.G. 1.183)



Data and Assumptions 7/171 SAs severely damaged Source term Fission Products: Kr, Xe, I, Sr, Cs, 3H ENEA

Activation: 3H (CR), 210Po (Pb) ENEA

100% release from pins & immediate release Volatilization fractions 800ºC (ULOF) I, Cs, Sr, Po (by KTH)

Radionuclides Noble Gases Iodine Caesium Strontium Polonium Tritium

Volatilization Fraction(800ºC - ULOF)

100% 2.99E-5 4.91E-6 5.34E-13 2.96E-7 100%

Volatilization Fraction(480ºC)

100% 9.51E-8 2.41E-7 7.53E-16 2.09E-10 100%



Data and Assumptions Containment Model

Chemical form of Iodine isotopes R.G. 1.183 Irrelevant (No removal) Radionuclide removal from containment only decay and leakage, NO

natural deposition, NO spray Primary Containment Leak rate 0.1% d-1 (EUR for LWR, Vol. 2, Rev. C)

Dose Conversion Factors FGR-11&12

X/Q need of meteorological data Pitesti (Romania) Reference values for accident condition (Spanish NPP)

SiteDistance

(m)

X/Q (s/m3)

0-2 h 2-8 h 8-24 h 24-96 h 96 h – end

Spanish NPP 1000 3.75E-4 1.91E-4 1.36E-4 6.66E-5 2.40E-5



Model for RADTRAD calculation:

Gap Release Coolant Release Cover G. Release

Fraction gap→Pb Dissol. Pb→CG Fraction CG→Cnt. Spray Depos. Decay Leakage Filtering Decay

Noble Gas 100% Instantan. NO Instantan. 1.00E+00 Instantan. NO NO YES YES NO NO

Iodine 100% Instantan. YES Instantan. 2.99E-05 Instantan. NO NO YES YES NO NO

Cesium 100% Instantan. YES Instantan. 4.91E-06 Instantan. NO NO YES YES NO NO

Strontium 100% Instantan. YES Instantan. 5.34E-13 Instantan. NO NO YES YES NO NO3H (TF) 100% Instantan. NO Instantan. 1.00E+00 Instantan. NO NO YES YES NO NO

3H (Act.) - N/A NO Instantan. 1.00E+00 Instantan. NO NO YES YES NO NO

Polonium - N/A YES Instantan. 2.96E-07 Instantan. NO NO YES YES NO NO

Removal after releaseFamily Isotope

Containment Removal (t: 0-720h)

Model simplifications

Release into containment = 100% Core Inventory (instantaneoust=0)

x [Volatilization fraction]

Containment Removal: only DECAY and LEAKAGE (environment)

NO Sprays, NO Deposition



PRIMARY COOLANT

(LEAD)

100% Release fromfuel & No

Dissolution

100% Release fromfuel & Dissolution

in coolant

PRIMARY COOLANT

(LEAD)

100% Release fromfuel & No

Dissolution

100% Release fromfuel & Dissolution

in coolant

NG (Xe, Kr)

I, Cs, Sr (1)

CONTAINMENT ATMOSPHERE

100% Release

CalculatedRelease Fraction

(KTH)

CalculatedRelease Fraction

(KTH) (2)

210Po

7 damagedSA’s

7 damagedSA’s

Activation of Pb andBi impurities

Activation of Pb andBi impurities

ENVIRONMENT

Leakage: 0.1% vol/day

Leakage: 0.1% vol/day

RELEASE MODEL

(2) There are Hg, Tl, Bi, Zn, Ag, In, Cd…isotopes, not covered by this

assessment

(1) There are other FPs, not covered by this

assessment

& Tritium(including

activation of B-10)

CORE

Model for RADTRAD calculation:



Activity Released to Environment:Gases Activity released to the environment

1.0E+10

1.0E+11

1.0E+12

1.0E+13

1.0E+14

1.0E+15

1 10 100 1000t (h)

A (B

q)

Xe-133Xe-133mXe-135Kr-85mKr-85Kr-87Kr-88H-3

I, Cs, Sr and Po Activity released to the environment

1.0E+04

1.0E+05

1.0E+06

1.0E+07

1.0E+08

1.0E+09

1.0E+10

1 10 100 1000t (h)

A (B

q)I-131I-132I-133I-134I-135Cs-134Cs-137Sr-90Po-210

True Gases

Lead-soluble FP & AP

~1012-1014 Bq (~100%)

~107-1010 Bq (<0.01%)



Whole-body dose

Xenon & Krypton~0.1 mSv after 30 days (~100%)

1.0E-05

1.0E-04

1.0E-03

1.0E-02

1.0E-01

0 100 200 300 400 500 600 700 800

t (h)

Dac

(mSv

)

Kr-85mKr-87Kr-88Xe-133mXe-133Xe-135mXe-135Xe-138Total

88Kr+135Xe+133Xe (84%)



Thyroid dose

Iodine & Tritium~0.03 mSv after 30 days (~100%) 131I+3H+133I (99%)

1.0E-07

1.0E-06

1.0E-05

1.0E-04

1.0E-03

1.0E-02

1.0E-01

0 100 200 300 400 500 600 700 800

t (h)

Dac

(mSv

)

I-131I-132I-133I-135H-3Po-210Cs-137Total



TEDE

Noble Gases & Tritium~0.1 mSv after 30 days (100%)

Xe-135m, 1%

H-3, 13%

Xe-135, 29%

I-131, 0%I-132, 0%

Xe-133, 16%

Kr-88, 28%

Cs-137, 0%

Cs-135, 0%

Cs-134, 0%

Xe-138, 6%

Xe-133m, 0%Xe-131m, 0%I-135, 0%

I-134, 0%

I-133, 0%

Kr-85m, 2% Kr-83m, 0%

Kr-87, 5%

Kr-85, 0%

ALFRED (LEADER) (J anuary 2013)



Doses

Whole-body, Thyroid & TEDE~0.04-0.001% of US NRC limits after 30 days

CalculationWhole-body

(mSv)Thyroid(mSv)

TEDE(mSv)

NG, Sr, Cs, I 9.43E-02 1.58E-02 9.48E-02

3H 1.16E-06 1.42E-02 1.42E-02

210Po 1.93E-13 4.40E-05 4.37E-04

Total 9.43E-02 3.00E-02 1.09E-01

NRC Limit250

(R.G. 1.195)3000

(R.G. 1.195)

250(R.G.1.183)Target: 50

% Limit 0.038 0.0010.044

Target: 0.22



Doses – Sensitivity Analysis

Whole-body, Thyroid & TEDEX/Q*10 & leakage*10 Source term 100 times higher

CalculationWhole-body

(mSv)Thyroid(mSv)

TEDE(mSv)

NG, Sr, Cs, I 9.36E+00 1.52E+00 9.41E+00

3H 1.05E-04 1.30E+00 1.30E+00

210Po 1.76E-11 4.02E-03 3.99E-02

Total 9.36E+00 2.82E+00 1.07E+01

NRC Limit250

(R.G. 1.195)3000

(R.G. 1.195)

250(R.G.1.183)Target: 50

% Limit 3.7 0.14.3

Target: 21.5



Conclusions

Compliance with present European Licensing Criteria

Conservative assumptions (damage, release, timing…)

No credit to engineered safeguards* (*Leak rate in accordance with non-loss of containment integrity)

~100% of whole-body dose due to Noble Gases

Thyroid dose due to I-131 and H-3 (No retention of tritium in lead)

~13% of TEDE due to Tritium (NO retention in lead)

Target leak rate: 0.1%/d radiologically adequate

Need of on-site meteorological data (X/Q)

Impact of dose methodology not expected to be high (Noble Gases)



D18 - Report on Containment Assessment for the ETDR Author: EA

Dates: Rev. 0 – January 2013 (status: for comments)

Rev. 1 – March 2013



Thank you!

Documents

WP 5 Safety and Transient Analysis Task 5.6 Containment and source term assessment for the ETDR LEADER Lead-cooled European Advanced DEmonstration Reactor