properties and behaviour of irradiated fuel under accident conditions - rondinella

1

Properties and behaviour of irradiated fuel

under accident conditions

V.V. Rondinella, R.J.M. Konings, J.-P. Glatz, P.D.W. Bottomley,

T.A.G. Wiss, D. Papaioannou, O. Benes, J.-Y. Colle, C.T. Walker,

S. Bremier, D. Serrano-Purroy, D. Staicu, D. Manara, L. Vlahovic,

P.Pöml, Th. Fanghänel

European Commission,

Joint Research Centre,

Institute for Transuranium Elements

P.O. Box 2340, 76125 Karlsruhe, Germany

http://itu.jrc.ec.europa.eu

[email protected]

IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012

IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 2

Outline

…Spent Fuel Safety in the Light of the Accident at the Fukushima

Daiichi Nuclear Power Plant

• context: safety of nuclear fuels and cycles at JRC-ITU

• previous studies on fuel under extreme/accident conditions

• refocusing activities

- source term: high T properties and behaviour

- source term: spent fuel corrosion in water

- spent fuel: impact load resistance and storage

• conclusions and perspectives


JRC: 7 research institutes in 5 EU countries

~2500 staff / 300 M€/a budget / 40 M€ income

Nuclear programme within the JRC

Nuclear Data, Reference Materials and Measurements

Fundamental Properties of Nuclear Materials and Applications

Waste Management and Environment

Reactors Safety

Fuels and Fuel Cycles Safety

Safeguards, Non-proliferation & Security

1957 European Atomic

Energy Community

(EURATOM)

Joint Research Centre (JRC)


NUCLEAR SAFETY and NUCLEAR SECURITY

Safety of

nuclear fuel

cycle / Nuclear

waste /

Environment

Exploratory/Discovery Research

Reference Centre for

policy makers, stakeholders and citizens

in the nuclear field

Training &

Education

Basic science,

Fundamental

properties &

Applications

Nuclear

safeguards,

Non-

Proliferation

&

security

The mission of JRC-ITU is to provide the scientific foundation for the protection of the European

citizen against risks associated with the handling and storage of highly radioactive material

Institute for Transuranium Elements


- samples synthesis, materials science studies

- PIE: safety during irradiation, (severe) accidents

- back-end: storage, disposal, P&T

- predictive tools: TRANSURANUS, multi-scale

fundamental approach

Conventional, Advanced Nuclear Fuels and Cycles

From basic actinide science,

to atomistic mechanisms

to operational fuel properties

Safety of nuclear fuel cycle at ITU

LWR fuel experience is basis

for studies on advanced fuels


samples synthesis (MA lab)

optical, acoustic microscopy; SEM

EPMA, SIMS; TEM-SEM; XRD

th. conductivity: laserflash, POLARIS

high T laser-heating (melting, vaporization, conductivity, high-P)

high T effusion, revaporization, annealing, KüFA (HTR)

non destructive rod examination: profilometry, radiography,

outer oxide layer, g-spectrometry

clad: H2-hydrides, creep, burst

(hot) indentation, impact-fracture

fission gas release, density

chemical analysis, laser ablation

separation (aqueous, pyro-)

leaching, electrochemistry

LWR

advanced reactors

HTR

high burnup

U, Th MOX

non-oxides

minor actinides

cladding/coating

Nuclear fuel studies at ITU

multidisciplinary approach

normal/off-normal operation

extreme conditions

storage

analytical/modeling tools

Competences Experimental tools

thermodynamics (Cp,

vapor pressure, melting point)

thermal transport

fission products, gases, minor

actinides:

phase distribution,

matter transport

radiation damage:

mechanisms and effects

microstructure – macroscopic

properties evolution

corrosion, creep

fuel restructuring

Scope (fuels)


Severe accidents programmes/networks: TMI,

Phebus, CIT, Coloss, SARNET II irradiated

fuel from real and simulated accidents

High T behaviour volatiles, fission gas

release; vapour pressure up to complete fuel

matrix vaporization; thermophysical properties

Basic thermodynamic data (Tm, phase

diagrams) actinides/fuel compounds, corium

and other systems

Spent fuel rod safety during storage and

transport mechanical stability 0 50 100

PuO2 (mol%)

2400

2600

2800

3000

3200

T/K

So lid so lu tio n

Liq u id so lu tio n

U-Pu oxide system

From conventional to advanced fuels safety UO2, MOX, Th-MOX, HTR (Küfa),

metal alloy, minor actinides, inert matrices, emerging concepts

Fuel under extreme conditions


Three Mile Island (TMI-2). A real accident; integral test (with incomplete data).

OECD-NEA led consortium under the initiative of US-DoE (INL) involving AEA, AECL, ANL,

CEA, KIT, JAEA, JRC-ITU, PSI, Studsvik.

Phebus FP test. Irradiated fuel bundle degradation and melting (1988-2012).

Integral test with good data collection, but still difficulties in interpretation. Led by IRSN

(France) and supported by the European Commission. USA, Canada, Japan, Korea and

Switzerland also participated.

Five integral tests under different conditions. On-line monitoring of bundle degradation, fp

release and subsequent behaviour in the simulated primary circuit and containment.

Corium Interaction Thermochemistry. EC Framework project, 8 partners (1997-1999)

small scale tests of liquid Zry dissolution of (irradiated) UO2, and modelling

Core Loss of geometry. High burnup UO2, MOX high T interaction with cladding

Revaporisation testing. EC Framework project, 3 partners

single effect tests of (re)volatilisation of fission products under different atmospheres

Severe accidents projects (selection)


Core bore rock G12-P9-B- (1000x, BSE);

phase density variations fully molten rock

U-rich Zr-rich ferrous (Fe, Ni, Cr)

core bore rocks

G12-P2-E, G12-P6-E,

G12-P9-B, G12-P10-A

upper

crust

D8-

P2,3

debris

samples

H8 7.2-7.9

fuel rod

remnant

C7 3-35

lower

crust

N5-P1-E

O7-P4

Debris H8-7-5-1 (40x); white pieces are UO2; long

grey piece is zircaloy-UO2 mix; banded structure

is zircaloy interacted with steel, Ni-based alloys

TMI-2


BSE SIMS 235U enrichment: in zircon 1.08%; in UOx inclusions 0.8%

Si Zr

Zircon crystal from Chernobyl “lava”

Pöml, Burakov et al., 2011

U

EPMA

Zircon: 3.1–14.6 wt.% UO2 (natural <1.5), Pu traces

UOx: ≈ 0.3 – 0.4 wt.% PuO2, Zr traces


PIE of Phebus bundle

FPT2 Disc 2 (lower surface) +51mm

Central rod missing – melt on north side

Zone with melt

good correspondence between tomography and sectional macrographs

ITU contribution:

- sectioning of degraded bundle into

14x2 cm discs

- microscopic examination and

analysis at selected points of the

bundle to establish the principal

interactions

- examination of PTA samples (filters)


Wood’s metal

CoriumDegraded fuel rod 7 with

cladding broken away

Molten materials with

filigree structure

Fully oxidised

cladding

Metallic

melt

Microscopic sample extracted by coring from Disc 2, FPT2, on lower surface at

+51.5mm BFC & its position in the disc tomography

PIE of Phebus bundle-coring


• Corium molten pool forms in a predictable geometry. Composition ~(U,Zr)O2. Rapid cool-

down leaves corium as a single, deformed cubic phase, slower cooling results in formation of

separate U-rich & Zr-rich oxides

• Samples reveal how Ag-Zr and Ni- (or Fe)-Zr interactions can create liquefied cladding

already by 1200°C (over 1700°C below UO2 melting) which can rapidly attack the fuel

• Irradiated fuel undergoes a more rapid degradation than non-irradiated fuel, because

-it is mechanically weak (pre-existing cracks)

-fg release & precipitation into bubbles lead to very high porosity: 'foaming' at very high T

-increased surface area for attack by corium

• Cs release <100%, some Cs remains in the overheated fuel and even in the melt pool

• Cs condenses on cooler surfaces (<700°C), but can easily revolatilise above 500°C in steam

(also in inert atmospheres) probably as CsOH - regardless of the deposit composition

Outcome

What type of information comes out of these studies

• mechanistic: mechanisms, rate

• thermodynamic: temperature, oxygen potentials

• thresholds: key materials, specific interactions & transition T (eg. Tm)


…Spent Fuel Safety in the Light of the Accident at the Fukushima

Daiichi Nuclear Power Plant

characterization of molten fuel/corium extended to cover specific

aspects relevant for the Fukushima analysis and remediation

refocusing activities

- source term: high T properties and behaviour

- source term: spent fuel corrosion in water

- spent fuel: impact load resistance and storage


0

0.2

0.4

0.6

0.8

1

1.2

800 1000 1200 1400 1600 1800 2000 2200 2400 2600 2800

Temperature (K)

No

rmali

zed

Fra

cti

on

al

Rele

ase

4He

86kr

96ZrO

129I

130Te

136Xe

137Cs

138Ba

139La

140Ce

139LaO

140CeO

238UO

88Sr

239PuO

87Rb

153Eu

150Sm

Source term studies: thermal release

central pellet region

Normalized fractional release of ~70 GWd/t UO2

4

2

3

5

5

6

8

9

7

10

14

14

12

To PRIMARYVACUUM13

11

MS, RANGE 1-500 AMU

TEMP.

Gamma, BataCounts

1

1

10

0.1-0.8mm

Al2O3

TC Gas inlet

SAMPLE

0.1-0.8mm

W

SAMPLE

15

steps

Ramp10-50K/min

To Q-GAMES(Quantitative GasMeasurementsystem)

TIME

Knudsen cell for effusion tests on irradiated fuel


Combined Knudsen cell – SEM analysis

outer surface fracture surface

in vacuum

preoxidized

500 750 1000 1250 1500 1750 2000

1E-13

1E-12

1E-11

1E-10

2.02.22.42.62.8

(b)

Td<-U3O8>

ma

ss s

ign

al (A

)

Temperature (K)

Cs

BaO

SrO

UOx tot

U3 O

8 UO

2

Td<U4O9>

(a)O/U

500 1000 1500 2000 2500 3000

1E-13

1E-12

1E-11

1E-10

1E-8

0.8

1.2

1.6

2.0

2.4

Sam

ple

com

ple

tely

vaporis

ed

Re

lea

se

qu

an

tity

(kg

/s)

Temperature (K)

90

Sr

129

I

130

Te

137

Cs

BaO

NdO

UO2

O/M

(a)

(b)

Source term: oxidation effects

morphology of ~65 GWd/t UO2 annealed at 1900 K

preoxidized

in vacuum

effusion behaviour

Hiernaut et al., 2008


Fragments (A, Filtered)

0,01

0,1

1

10

100

1000

0 25 50 75 100

Time (days)

FR

NU

Rb85

Cs133

Mo98

Zr/Sr90

Np237

Y89

Rh103

Ba138

Nd144

Pu240

U238

Ru102

Tc99

Zr93

Pd105

FRNU>1: Mo, Cs, Rb, Ba, Tc, Np, Sr(Zr)90

FRNU≈1: Y, Nd, (Np), Pu (≤1)

FRNU<1: Ru, Pd, Zr, Pu (≤1)

Fractional Release Normalized to U;

leaching of 35 GWd/t MOX in groundwater

Source term: water corrosion

UO2

UO2-0,1%238Pu

UO2-10%238Pu Studtite

Schoepite

Secondary phases on leached UO2

after Fukushima: spent fuel leaching in seawater ongoing

groundwater

leaching tests


a b c d

corrosion layer

removal zone

fuel release <2 g/break i.e. less than the mass of

a single fuel pellet

simulated impact

accident during spent

fuel rods transportation

safety of spent fuel

storage/transport

Impact load response of a ~74 GWd/t PWR rod (high speed camera sequence)

GNS-AREVA

collaboration D. Papaioannou et al., 2009

PWR and BWR rods

tested: 19 – 74 GWd/t

Fuel safety out of pile


Spent fuel storage

• No direct spent fuel data in the extended

range; decay damage saturation?

swelling? mechanical integrity?

• He accumulation during storage

exceeds solubility level. Will it all

accommodate in defects, fg bubbles?

Ongoing work to elucidate conditions

and mechanisms relevant for storage:

- spent fuel swelling/pressurization;

response to long term (low) T history

- cladding properties evolution

- microstructure alterations at high dose

- fuel composition/irradiation history effect

-decay and He production in spent fuel

Time from discharge, years

100 101 102 103 104 105 106

He

prod

uced

per

g o

f fu

el, g

10-7

10-6

10-5

10-4

10-3

10-2

-d

ecay

s g

-1

1017

1018

1019

1020

1021

UO2 40 GWd/tM

UO2 60GWd/tM

UO2 80 GWd/tM

UO2 100 GWd/tM

MOX 25 GWd/tM

MOX 45 GWd/tM

MOX 60 GWd/tM

~1 dpa

fuel 4

5% P

u

~0.01 dpa

eol

He solubility

~10 dpa

~100 dpa

Approach:

- spent fuel characterization

- accelerated -decay, He accumulation

- He solubility, thermodynamic equilibrium


Conclusions and perspectives

• Significant amount of knowledge on fuel behaviour during severe accidents exists

from international projects on fuel from actual or simulated severe accidents

• New programmes are proposed to extend the experimental basis of data for

modeling tools and fill some gaps. This will benefit from advances in experimental

characterization tools

• In JRC-ITU, some R&D activities on fuel safety are refocused to cover specific

issues related to the Fukushima accident and to its aftermath, e.g. molten

fuel/corium properties, source term assessment for high T release, corrosion effects

in seawater/salt, spent fuel behaviour in the pools, storage/treatment of molten fuel,

etc. Links/collaboration with Japanese partners (CRIEPI, JAEA) are being developed

• Integrated approaches are mandatory to optimize use of resources (less money

and time than in the past) and to investigate all systems

international partnerships/programmes

integrated experimental/theoretical


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properties and behaviour of irradiated fuel under accident conditions - rondinella

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