State of the Art for Severe Accident Source Term...

State of the Art for Severe Accident Source Term Estimation

Harri Tuomisto

Fortum Power, Finland

IAEA Technical Meeting on Source Term Evaluation for Severe Accidents,

Vienna, Austria, 21-23 October 2013

Contents

• Severe Accident Source Term: definitions

• Historical context: Concept of Maximum Credible Accident

• Concept of severe accidents

• Basic steps for source term estimation

– Core inventories

– Fission product releases

– Source terms into the containment

– In-vessel containment source term

• Radiological consequences

• Severe Accident Management for mitigation of radiological consequences

• State-of-the art reports related to severe accident source terms

• TECDOC – Source Term Evaluation in Severe Accidents

21 October 2013 Harri Tuomisto2

Severe Accident Source Term: definitions

• Source term into the containment is the magnitude, physical and chemical form

and timing of the release of fission products and other aerosols from core

materials and concrete to the primary containment atmosphere or to the

suppression pool from both in- and ex-vessel sources.

• In-containment source term is the airborne radioactivity and its physical and

chemical form in the atmosphere of the primary containment as a function of time.

Thus, the in-containment source term is the radioactivity that is available to be

released from the primary containment.

• In this presentation, the severe accident source term is defined as the release

of radioactive substances from the containment into the environment, which can

take place into the atmosphere or into the soil, as the consequence of a severe

accident.


Concept of severe accidents

• Nuclear power plant accidents are called severe accidents, when

they lead to extensive degradation of the reactor core.

• Severe accidents can progress to partial or full melting of the whole

reactor core.

• The core degradation resulting from reactivity initiated accidents can

be very destructive − they are not treated in this presentation with

the exception of the references made to the Chernobyl disaster

• Management of severe accidents is defined as Level 4 of the

defense-in-depth concept

21 October 2013 Harri Tuomisto

Historical context: Concept of Maximum Credible Accident (MCA)

• Deterministic approach, developed at the time, when knowledge was not sufficient to design the containment against severe accident phenomena

• MCA led to definition of Design Basis Accident (DBA) concept: – Large Break Loss-of-Coolant Accident (LBLOCA) was chosen to represent

thermal-hydraulic loading to containment and dimensioning of emergency cooling capacity

– core degradation was chosen as a basis for releases of radioactivities from the core and primary circuit (source term estimations based onTID-14844published in 1962)

– certain events were screened out such as reactor vessel failure, reactivity accidents leading to severe accident (probabilistic feature of deterministic approach)

21 October 2013 Harri Tuomisto

Historical context: Concept of Maximum Credible Accident (MCA)

• MCA concept worked well at Three Mile Island - but LBLOCA didn't work for core cooling (as PRA in WASH-1400 had already predicted this)

– Led to the development of more mechanistic understanding of reactor accident source terms that culminated in the Source Term Code Package and NUREG-1150 (Severe

Accident Risks: An Assessment for Five U.S. Nuclear Power Plants) and NUREG-1465(Accident Source Terms for Light-Water Nuclear Power Plants)

– Industry response through the IDCOR programme

• Screening of reactivity accidents failed at Chernobyl

– Led to reinforcing the Defence-in-Depth concept with introduction of Severe Accident

Management (SAM) and safety culture

• Failure to apply PSA to external hazards (failure of appreciation of

historical data for external hazards) at Fukushima

– Led to in the first phase to the “stress tests”, reinforcing SAM and mobile equipment.


Progress of severe accidents

• Severe accidents initiate, when reactor core cooling can't be restored after a

transient or accident

• If reactor circuit is intact, operators try to restore core cooling by preventive

SAM measures e.g. by inititiating bleed and feed action in the secondary

circuit, and if not successful then in the reactor circuit

• If reactor circuit leaks, operators try to inject coolant to primary circuit by any

available means (in case of PWR, however, borated water is needed)

• If bleed and feed actions are not successful, such sequences lead to

uncovery and overcooling of the reactor core

• In case of no cooling, reactor core eventually degrades and melts and

relocates on the reaactor vessel lower head (molten core materials are

referred as 'corium')


Progress of severe accidents

• If the core melt progression can be stabilized on the lower head (e.g. byexternal cooling of the vessel), the ex-vessel (=in-containment) consequences are less severe

• However, fission products, hydrogen and decay heat are released in largeamounts to the containment atmosphere

• If corium melts through the vessel, there are various energetic consequencescaused by ejected high-temperature molten corium

• Molten corium slumping to water pools (either in-vessel or ex-vessel), orpouring water on the molten material surface, may cause energetic steamexplosions

• Molten corium on the containment basemat initiates core-concrete interactionthat releases aerosols and non-condensible gases to containmentatmosphere and erodes the concrete ('China syndrome')

• Overpressure formation in the containment


Harri Tuomisto 9

Basic steps for source term estimation

1. Estimate the inventory of fission products in the core.

2. Estimate the amount of fission product release from the core.

3. Estimate the source term into the containment

– Identify the release pathways,

– Identify and characterize the dominant transport phenomena,

– Identify and estimate the ex-vessel release rates

4. Estimate the in-containment source term

– Identify and estimate the impact of the retention mechanisms in the containment

5. Estimate the releases of radioactive substances into the environment (severe accident source term)

21 October 2013

Fission product yield and inventory

In order to predict reliably the fission

product inventory in irradiated fuel, it is

essential to know the neutron induced

fission yields.

The fission product inventory of the

reactor core depends mainly on the fuel

composition, amount and burn-up.

The fission product inventories in the fuel

are calculated using separate codes such

as ORIGEN and FISPIN.


Harri Tuomisto

Radionuclide Groups & Typical Inventory

Source: AP-600

End-of-Cycle Mass in Core

(kg) Group

No.

Name (representative

element) Elements Contained in Group

PWR

1 Noble gases Xe, Kr 412

2 Iodine I, Br 18

3 Cesium Cs, Rb 238

4 Tellurium Te, Sb, Se 34

5 Strontium Sr 71

6 Ruthenium Ru, Rh, Pd, Mo, Tc 612

7 Lanthanum La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y 567

8 Cerium Ce, Pu, Np 201

9 Barium Ba 108

Fission product release

• In LWRs, the timing of four accident phases is treated separately:

– Gap release (clad ballooning and rupture)

– In-vessel release (core degradation, molten pool formation)

– Ex-vessel release (MCCI etc.)

– Late in-vessel release (in-vessel retention, revaporization)


Harri Tuomisto

Case: PWR releases into containment during successive phases− for illustration purpose

Gap Release Early In-Vessel Ex-Vessel Late In-Vessel

Duration (hours) 0.5 1.3 2.0 10.0

Noble Gases 0.05 0.95 0 0

Halogens 0.05 0.35 0.25 0.1

Alkali Metals 0.05 0.25 0.35 0.1

Tellurium group 0 0.05 0.25 0.005

Barium, Strontium 0 0.02 0.1 0

Noble Metals 0 0.0025 0.0025 0

Cerium group 0 0.0005 0.005 0

Lanthanides 0 0.0002 0.005 0

Values shown are fractions of core inventory

21 October 2013

Fission product release is a strong function of temperature

gap release of volatiles release of release of refractory

release semi-volatiles metals / ceramics

Xe, Kr I, Cs Te Sr, Ba Ru, La, Ce

Zr steel Eutectic fuel (UO2) melting

oxidation melting dissolution

clad core heatup, degradation, and relocation core-concrete interactions

failure

1000 1400 1800 2200 2600 3000

Temperature (C)

Fission products can be divided to the groups according their volatility:• Noble gases (Kr, Xe)• Very volatile (I, Cs)

• Moderately volatile (Te, Sr, Ba)• Less volatile (Ru, La, Ce)


Fission product release occurs at different times


Fission product transport in the RCS

• Fission products (+actinides and activated structural materials) are released as gases or vapors from the degrading core.

• Aerosol formation and agglomeration: the aerosols are swept by a steam-hydrogen mixture towards the breach

• A number of important physico-chemical processes occur on the way.

• These processes reduce the quantity of material released into the containment, and they also condition its physico-chemical form.

• Transport times are short due to the short distance and the high flow rates.

• There are various aerosol retention mechanisms taking place: Brownian diffusion,

thermophoresis, diffusiophoresis, electrophoresis, sedimentation (gravitational

settling), inertial impaction and pool scrubbing.

• Particles deposited on the surfaces of the primary coolant circuit may resuspendback to the gas stream (leads to the decrease of retention into the circuit).

• Resuspension is especially important in bypass sequences, in which the

radionuclides may be released directly to the environment.


Harri Tuomisto 17

Aerosol behaviour considered in containment analyses

• Aerosol behaviour in the containment is complex and sequence-specific.

• Deposition and resuspension are

significant sources of uncertainties in case of late containment failure (few

hours after the release from the RCS).

• Deposition by gravitation is the most important retention mechanism with no

sprays. Particle growth is an important uncertainty affecting the removal rate.

• In modelling the important parameters are relative humidity and temperature gradients at structure surfaces.

• Hygroscopic or soluble aerosols (even in superheated conditions): faster particle

growth and deposition inside the containment.

21 October 2013

Iodine chemistry

All feasible main interrelation and feedback processes between iodine chemistry, thermal hydraulics and aerosol physics in a LWR containment:

Simplified diagram of iodine transformations within the containment:

Extracted figures from: OECD NEA/CSNI/R(2007)1 State-of-the Art Report on Iodine Chemistry


Aerosol removal processes

Aerosol removal processes have a significant impact on the predicted environmental

source term associated with late containment failure.

– removal by containment sprays,

– removing iodine by chemical controlling e.g. of sprays,

– pool scrubbing:

• overlying water pool to prevent radionuclides becoming airborne,

• filtering the release path

– using filtration to reduce the source term during vented and intact containment sequences,

– aerosol deposition/plugging in narrow leakage paths (intact containment sequences).


Radiological consequences

• The IAEA Fundamental Safety Principles

– Measures for controlling radiation risks must ensure that no individual bears an

unacceptable risk of harm, and that people and the environment must be

protected against radiation risks.

• For assessment of the radiation risks, radiological consequences of

the NPP states should be predicted.

• Radiological consequences mean on site and off-site effects of

ionizing radiation on people and environment resulting from different

plant states, including severe accidents.


Criteria for radiological consequences

• Mitigation of consequences of severe accidents is usually formulated through

probabilistic risk criteria: Defining an upper frequency limit for severe accident

sequences resulting in unacceptable consequences.

• However, in several countries the high level (deterministic) radiological criteria

have been specified also for beyond design basis accidents and severe

accidents.

• These consequences can be defined in terms of released radioactivities and

effective dose to critical groups.

• Criteria are intended to ensure that there will be no short-term or long-term

deterministic health effects or no long-term restriction of use of large land or

water areas.

• Short-term criteria are often related to releases of iodine I-131 and long-term

criteria to releases of caesium Cs-137) or groups of radioisotopes.


Severe Accident Management for mitigation of radiological consequences

Effective SAM measures in mitigating

the radiological consequences:

• Level II: minimize possibilities of the

impaired containment function

• Level III: ensure containment

isolation

• Level IV: ensure that there are no

catastrophic containment failures

(hydrogen management, filtered

venting etc)

• Level V: stabilization of the core

melt (end of release)


Passive systems having impact on source terms

• Containment designs, particularly double containments

• Core catchers

• In-vessel retention (IVR)

• Passive autocatalytic recombiners (PAR)

• Passive containment cooling systems

IVR has been applied in some existing plants.

PARs have been installed in many existing plants.

Core catchers and passive containment cooling systems are typical design

features of Generation III+ reactors.


State-of-the art reports related to severe accident source terms

• NUREG-1465 (1995): Accident Source

Terms for Light Water Nuclear Power

Plants

• OECD NEA/CSNI/R(2009)5 State-of-

the Art Report on Nuclear Aerosols

• OECD NEA/CSNI/R(2007)1 State-of-

the Art Report on Iodine Chemistry

• IAEA, Approaches and Tools for Severe

Accident Analysis for Nuclear Power

Plants, Safety Report No. 56, 2008


TECDOC – Source Term Evaluation in Severe Accidents

The proposed TECDOC with the following contents:

1. Introduction

2. Design and regulatory use of source terms

3. Severe accident sequences

4. Release of fission products

5. Fission product removal from the containment atmosphere

6. Fission product retention in Generation III reactors

7. Radiological consequences of severe accidents

This Technical meeting is expected to provide

• insights and input for a SOA TECDOC,

• emerging issues in STE following the Fukushima Daiichi accident.


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