astec code fission product release – models and evaluation … · astec code fission product...

presented by Nils Reinke

GRS Cologne

ASTEC Code

Fission Product Release

– Models and Evaluation –

IAEA – Technical Meeting on

Source Term Evaluation for Severe Accidents

Vienna, Austria, 21 - 23 October 2013

Presentation includes information from IRSN and GRS

IAEA – TM – Source Term Evaluation for Severe Accidents, Oct. 22, 2013 2

� ASTEC Basics

� FP Release & Behavior

� ASTEC Validation

� Conclusions & Outlook

ASTEC – web site

ASTEC – an IRSN / GRS Co-operation

Content of the presentation

ASTEC – Basics


� ASTEC (Accident Source Term Evaluation Code) developed by

IRSN (France) and GRS (Germany)

� for LWR (present/future PWRs, BWRs, VVERs) source term severe accident calculations,

from initiating events until radioactive releases out of the containment

� validated versus many experiments and OECD / NEA ISP exercises over more than 15 ys

� Main requirements:

• fast-running code i.e. < real time sequence

• accounting for safety systems and their availability (SAM)

• high level of model validation

• modular, flexible (applications), user-friendly

� Continued development within several European Commission’s co-sponsored projects:

ASTEC-V0

2003

V1.1

20091999

V1.0

2004

V2.0

MOU

1994

EC-VASA 1994-1998 EC-SARNET 04-08EC-EVITA 2000-03 EC-SARNET2 09-13

2013

V2.0

EC-CESAM 13-17

2014

V2.1

ASTEC – Basics


� Analyses of Severe Accidents, like

station black out, loss of steam generator

feed-water, steam generator tube

ruptures, as well as small, medium,

and large break loss of coolant

accidents

� Reference Datasets for generic

types of NPPs in Europe (PWR,

in the near future also BWR),

PHWR (and CANDU) are being

developed and qualified

� Plant Analyses are performed

related to:

• Accident Management and

Source Term determination studies

• Probabilistic Safety Assessment level 2 (PSA-2) studies

• Lessons learnt from the Fukushima accidents

ASTEC – Basics


Fission Product Release & Behavior – ASTEC Modules


� In-vessel fission product release

�� ELSA module

� In-vessel (coolant circuit, PC & SC)

fission product behavior

�� SOPHAEROS Module

� Ex-vessel fission product release

�� MEDICIS module

� Ex-vessel fission product behavior

incl. iodine

�� CPA & IODE module

� Semi-empirical and mechanistic

approaches (integral code demands)

� Must be compatible with the range of conditions to be covered and the level of

confidence required in the predictions


In-Vessel Fission Product Release

In-Vessel Fission Product Release – ELSA Module


� Fission Product distribution

within fuel rods

STRU FPEVOL

PHYS ‘ELSA’

SC1 EMIT ‘fuel1’ ‘fuel2’ TERM

SC1 BARR ‘clad1’ ‘clad2’ TERM

SR1 FACT 1. 1. TERM

! factor for radial distribution of fission products

SRG GAP ‘XE’ 1.D-3 ‘KR’ 1.D-3 TERM

SR1 PROF 0. 0.9

1. 1.0

2. 1.1

3. 1.0

4. 0.9 TERM

! axial profile giving the FP distribution along fuel rods

END

radial profile

axial profile



� ASTEC approach in the ELSA module

• semi-empirical approach with 3 classes (volatile, semi-volatile and low-volatile

species), keeping only the limiting phenomena



� List of uncertain input parameters for ELSA model

• the ratio surface/volume of the fuel pellets due to roughness

• the ratio surface/volume of the fuel pellets for the limited steam access

• the ratio surface/volume of the debris due to roughness

• the surface exchange of the SIC release at the cladding failure

• the surface exchange of the SIC release during the candling

• Number of grain size classes for fuel pellets

• Lower bound of the grain size distribution, upper bound of the grain size

distribution, standard deviation, geometrical diameter

• Choice of the modeling for the stoichiometric deviation

� Recommended default values are advised for the reactor calculations

� SOPHAEROS predicts vapor and aerosols transport in the reactor coolant

system (RCS) formed by condensation of material released from the degraded

core

� important step in assessment of final radiological source term.

� FP vapors condense to form aerosol (volatility varies according to speciation)

� ^transport of aerosols/vapors with

flowing atmosphere

� aerosols/vapors can deposit

on surfaces

� transported decay heat

provides heat source

where FP are deposited

(RCS pipes and SG tubes).

� deposited aerosols can

revaporize if surface heats

sufficiently

FP Behavior in Reactor Coolant System – SOPHAEROS Module


FP Behavior in RCS – SOPHAEROS Module


� Aerosol models – phenomena taken into account:

• Agglomeration:

− Gravitational: Pruppacher & Klett,

− Brownian diffusion,

− Turbulence, shear and inertial: Saffman & Turner.

• Deposition:

− Gravitational: Stokes + Cunningham correction,

− Diffusion: Brownian (Gormley & Kennedy) or turbulent (Davies),

− Turbulent impaction: Liu & Argawal,

− Thermophoresis: Talbot et al.,

− Diffusiophoresis: Waldmann or Loyalka based on Stefan velocity,

− Bend impaction: centrifugal or Cheng & Wang/Pui et al.

• Remobilisation of deposits:

− Thermal re-vaporisation,

− Mechanical resuspension: drag-lift-adhesion model


Ex-Vessel Fission Product Release

FP Behavior in Containment – CPA Module


1- Gas

2- Aerosol

3- Gas phase surfaces

4- Water

5- Water phase surfaces

6- Safety systems (filter, etc.)

� Aerosol and FP behaviour:

• aerosol (insoluble, hygroscopic) transport and depletion models:

sedimentation, diffusiophoresis, thermophoresis

• condensation of aerosols,

• aerosol retention through water

pools (pool-scrubbing model

SPARC-B),

• aerosol removal by spray,

including wash-out of deposits

on containment walls,

• FP transport and calculation

of decay heat of gaseous and

particulate FP,

• FP/aerosols retention in concrete

walls cracks.

• filter models: granulate and fibre filter types,

retention due to: interception, impaction, diffusion

FP Behavior in Containment – IODE Module


� Chemical reactions in sump and gas phase

in each containment zone,

• reactions in liquid phase:

− hydrolysis of molecular iodine,

radiolytic oxidation of I-, pH influence …

• reactions in gas phase:

− oxidation of molecular iodine by O3 air

radiolysis products, formation of organic

iodine (CH3I) from painted walls,

− adsorbtion/desorption of molecular iodine on

walls,

− CH3I destruction, O3 formation,

• mass transfers sump-gas phase.

R

ADSORPTION/

DESORPTION OF I2

Organics release

AerosolsIode-MétalCsI, AgI, CdI2…

I2R

CH3IOrganics release

CH3I(g)I2(g)

settling

air + H2O �� O3

IxOyNz +

IxOy

I2

CH

I-, HOI, IO3-

3I

R

Adsorption/desorption of I2

Ex-Vessel Fission Product Release/behaviour – MCCI etc.


� Release during corium-concrete interaction (MCCI)

• similar to ELSA (in-vessel) approach

• chemical equilibrium in a well-mixed molten pool assumed

• only semi- and low-volatiles are concerned

• release can be evaluated by thermodynamic calculations

• release for Zr-rich corium and siliceous concrete

� release from boiling sump

• re-suspension of aerosol trapped in sump water

• semi-empirical models deduced from REST (FZ-Karlsruhe) experiments

• lower release as compared to in-vessel release (but long term process)

� mechanical re-suspension of aerosols deposited on containment walls / floors

(e.g. in case of hydrogen deflagration)


ASTEC validation

ASTEC Validation


Illustration of main experiments used for ASTEC validation for FP release/behavior

ASTEC Validation – VERCORS Experiments


� The VERCORS experimental program (CEA)

� radionuclide release from standard PWR fuels

� twenty-five experiments have been

performed leading to a large database

regarding release of fission products

from UO2 and MOX fuels

under several types of atmosphere.

� The fuel burn-up ranges from

38 GWd/t to 70 GWd/tU.



� VERCORS 4 (Fuel burn-up 38.3 GWd/tU)

� key parameters: the temperature ramp, the burn-up of the fuel sample, gas

composition (steam and/or hydrogen), and flow rate

� oxidizing plateau with a mixed steam and hydrogen flow at a temperature of

around 1570 K preceded most of these tests, in order to completely oxidize the

cladding before the last heating ramp to the final high temperature plateau

Temperature volatile

FP release



Low-volatile

FP release

Semi-volatile

FP release

ASTEC Validation – Phebus Experiments


� Irradiated fuel heated in test package by Phebus driver core

� objectives:

• Fuel heat-up

• Zr oxidation, H2 release

• Fission product behavior

� circuit (700°C) transports FP

through steam generator tube

• deposits in circuit and SG

� containment receives FP gas

and aerosols

• settling…

• Iodine chemistry

� involves most SA phenomena

� important validation task for all integral codes, in particular for source term

evaluation: FPT0, FPT1 (ISP46), FPT2, FPT3 (with B4C), FPT4.



In-vessel volatile FP release – ELSA results

FP release to containment –

SOPHAEROS results



ELSA results

In-vessel semi-volatile FP release In-vessel low-volatile FP release

annulus air

extraction system

environment

filter20OBKUPPEL

18KUPPELA

11OPERA

10UPERA

21RRUNTEN

22RRMITTE

23 RROBEN

12UPERB

5DH

HKPB

4HKPA

19KUPPELB

13OPERB

FILTER24

ENVIRON

16UKUPA

17UKUPB

8DEOBOXA

9DEOBOXB

7DEMBOXB

14RRAUM

CAVITY

3PKLB

1SUMPF

1SUMPF

2PKLA

6DEMBOXA

ATMOS_JU

RUPTURE

DRAIN_BOT

� Severe accident sequence for

German KONVOI 1300 MW PWR

� Loss-of-normal-feedwater (LOFW)

(end of calculation 150.000 s)

� comparison between

MELCOR and ASTEC

� feasibility study on

ASTEC application to

PSA level 2

� emphasis on accident

sequence modeling

� RPV failure at approx. 10 h

� no source term to

environment calculated

� but possible release path via

annulus air extraction system

ASTEC – Plant Analyses


annulus



� Source term for KONVOI PWR LOFW (at end of calculation 150.000 s)

� possible release from annulus via air extraction system (filtered) into environment

� orders-of-magnitute agreement despite different modeling approaches

MELCOR ASTEC MELCOR ASTEC MELCOR ASTEC MELCOR ASTEC MELCOR ASTEC

Xe 9,91E-01 9,99E-01 2,29E-05 1,03E-03 9,45E-04 3,32E-03 9,87E-01 9,96E-01 2,93E-03 3,99E-03

CsOH 9,85E-01 9,99E-01 3,34E-05 1,01E-03 1,18E-01 1,35E-01 8,67E-01 8,64E-01 1,14E-04 2,26E-05

Ba 4,20E-02 6,21E-03 9,43E-01 9,90E-01 4,29E-04 5,86E-03 4,15E-02 3,56E-04 3,37E-06 0,00E+00

Te 9,08E-01 9,75E-01 8,05E-02 2,52E-02 5,39E-02 2,78E-01 8,54E-01 6,97E-01 1,01E-04 1,70E-05

Ru 4,75E-02 1,47E-04 9,37E-01 9,58E-01 6,69E-03 1,02E-04 4,08E-02 4,52E-05 5,40E-06 0,00E+00

Mo als Element 1,22E-01 8,61E-01 2,27E-04 3,01E-02 1,35E-03 9,22E-02 1,73E-07 2,46E-06

Mo als Cs2MoO4 1,78E-03 5,06E-02 3,26E-06

Ce 3,47E-03 3,71E-05 9,81E-01 9,59E-01 3,86E-06 1,16E-05 3,46E-03 2,56E-05 1,26E-07 0,00E+00

La 3,62E-03 3,75E-05 9,81E-01 9,59E-01 3,19E-04 1,19E-05 3,30E-03 2,56E-05 4,19E-07 0,00E+00

U 5,10E-04 3,71E-05 9,76E-01 9,59E-01 6,96E-05 1,06E-05 4,40E-04 2,65E-05 5,74E-08 2,68E-08

Cd (Sb) 6,32E-01 9,82E-01 3,57E-01 1,76E-02 7,21E-02 1,31E-01 5,60E-01 8,52E-01 7,35E-05 0,00E+00

Sn 6,07E-01 3,84E-05 3,81E-01 9,59E-01 7,19E-02 2,79E-05 5,35E-01 1,05E-05 7,06E-05 0,00E+00

CsI (I) 9,66E-01 9,99E-01 2,34E-02 1,04E-03 4,06E-02 1,04E-03 9,25E-01 8,58E-01 1,03E-04 1,45E-05

Die Freisetzungsanteile wurden auf das Anfangsinventar der radioaktiven Spaltprodukte normiert

5,39E-02 9,46E-01

Gesamt aus Kern Verbleibend im Kern Im RKL Im Containment Außerhalb Containmentreleased from core remaining in core in primary circuit in annulusin containment



� Additional analyses performed by IRSN on source term

assuming:

• various initiating events : LOCA, LFW, SBO

• location and size of the break : hot and cold leg

• availabilities of safety systems

• that the containment and the containment venting system (filtered containment

venting (FCV)) won’t fail

� Preliminary results focusing on iodine:

• uncertainties on iodine phenomenology knowledge have an important impact,

in the same order of magnitude as the variability due to the scenario

• gaseous iodine mass fraction and iodine oxide mass deposition rate are the

major contributors to these uncertainties

• sensitivity analysis module (SUNSET) helps to establish ranking of the studied

effects

Conclusions – Severe Accident and Source Term Analyses


� Develop adequate input for codes

• requires high knowledge of code user on severe accident phenomena

• need for adequate and sufficient information on plant specifics and design

• use real plant data without conservative assumptions

• need for appropriate modelling of relevant plant specifics and all probable fission product release paths into the environment

• need for sufficient detail of nodalisation schemes for all components and buildings to allow a realistic simulation of NPP behaviour under severe accident conditions

� Validate developed input decks

• against real plant data for normal plant operating conditions

• by code to code comparisons with detailed (mechanistic) codes

� Perform uncertainty and sensitivity analyses

� Future ASTEC V2.1 version (end of 2014), which main features will be:

• generalization of the Material Data Base (MDB) use by all ASTEC modules

• new CESAR (T/H) / ICARE (core degradation) coupling

to account for e.g. late phase core re-flooding

• capability to simulate air ingress situations

incl. SFP behavior

• improved DCH modeling

• progress towards more complete BWR and PHWR (CANDU) applications

− description of canister components and account for multiple

in-core sub-channels (intra- and inter-canister flows),

− necessary adaptations to modeling core configuration with view to both

heat transfer and corium relocation models

− preliminary calculations of Fukushima accidents are already underway

Conclusions – Outlook



Thank you for your attention…

The author appreciates valuable contributions from the whole ASTEC team from IRSN and GRS

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