Upload
iaeainformation
View
195
Download
1
Embed Size (px)
DESCRIPTION
Wednesday, 21.03.2012, Spent Fuel Session
Citation preview
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
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
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 3
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)
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 4
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
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 5
- 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
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 6
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)
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 7
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
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 8
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)
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 9
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
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 10
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
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 11
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)
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 12
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
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 13
• 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)
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 14
…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
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 15
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
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 16
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
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 17
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
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 18
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
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 19
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
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 20
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
IAEA, International Experts’ Meeting, Vienna, 19-22 March 2012 21
23