Overview of Experimental Programs on Core
Melt Progression and Fission Product Release
Behaviour
B.J. Lewis, Royal Military College of Canada
R. Dickson, Atomic Energy of Canada Limited
F.C. Iglesias, Candesco Corporation
International VERCORS Seminar
Gréoux les bains, France
October 15-16, 2007
Outline
 Experiment Review
 Integral Severe Accident and Single Effect Tests
 Degraded Core Accident Phenomena
 Fission Product Release (FPR) Behavior
In-Pile Tests
 Source Term Experiments Project (STEP 1,2,3,4)
 Fission product release (FPR) and aerosol chemistry
 Source Term Tests (ST 1,2)
 FPR & aerosols from highly-irradiated fuel (reducing conditions)
 Damaged Fuel (DF 1,2,3,4) Relocation Experiment
 Coolant flow rate, system/fuel-rod pressure, degree of initial clad oxidation
 Severe Fuel Damage Tests (SFD ST, 1-1, 1-3, 1-4)
 Fuel bundle & FPR (transport/deposition) behavior, H2 generation
In-Pile Tests Cont’d
 Full Length High Temperature Tests (FLHT 1,2,4,5)
 Oxidation & H2 generation in full-length rods
 Loss-of-Fluid Test Facility Fission Product Test (LOFT FP-2)
 Large-scale test on FPR, steam supply/reflood
 Melt Progression (MP 1,2)
 Ceramic pool behavior in blocked-core accidents

Blowdown Test Facility (BTF-104, -105A, -105B, -107)
 CANDU fuel and FPR behaviour

Phebus SFD  Phebus Fission Product Tests (FPT-0,-1,-2,-3-4)
 Core, cooling system & containment response including FPR
transport/deposition
 Semi-volatile & actinide release from UO2/ZrO2 rubble bed
Out-of-Pile Tests (Electrically Heated)
 CORA (19 tests)  Quench
 Temporal behavior of core melt progression & reflood

PARAMETER (UO2 pellets and VVER cladding (1% Nb))
Bundle Configurations
Out-of-Pile Annealing Tests
 Single Effects Fission Product Release (FPR) Experiments:
 FPR from spent fuel (hydrogen, steam, air)
 ORNL: Horizontal Induction (HI 1-6), Vertical Induction (VI 1-7)
 CEA-CENG: HEVA 1-8, Vercors 1-6, High Temperature (HT 1-3), Release of
Transuranics (RT 1-8)
 AECL-CRL: >300 tests, e.g., MCE1-1,-6,-7, MCE2-13,-19, HCE2-BM3,CM4, UCE12-8
 JAERI: Verification Experiments of radionuclides Gas/Aerosol release (VEGA
1-10)
ORNL Experiments

Test conditions:
 Highly-irradiated Zircaloy-clad UO2 fuel samples 15-20 cm long (100-200 g)
 Atmospheric pressure up to 1700-2700 K (time at temperature 2-60 min)
 Test atmosphere: steam in VI-3, hydrogen in VI-5, hydrogen followed by steam in
VI-6, and air and steam in VI-7 (investigate atmospheric effect on FPR)

Major differences between VI and HI tests:
 VI tests (vertical) vs HI tests (horizontal)
 VI test conducted at higher burnup and temperature (2300-2700 K)

Measurements obtained
 Sample temperature vs time measured by optical pyrometry
 Thermal gradient tube (TGT) downstream to collect condensing vapors
 Graduated filters & impregnated charcoal cartridges to collect particulates and
volatile I species
 Charcoal cold trap to measure fission gases
 On-line measurements of fuel location and Cs-137 in TGT and Kr-85 in gas traps
 Post-test analysis of all components by gamma-ray spectrometry, NAA, sparksource mass spec, emission spectrometry
ORNL Experiments Cont’d

Major test results:
 Similar release rates for noble gases, Cs and I
 Difference Cs transport behavior in steam relative to hydrogen.
 Reactive vapor forms of Cs predominate in hydrogen conditions and transportable
aerosols in steam
 Similar Te and Sb release from UO2 as for volatile FPs but retained by metallic
Zircaloy until nearly complete clad oxidation
 Both Eu and Sb showed sensitivity to oxygen potential at high temperature.

Limitations:
 Segmented furnace tube did not provide good containment of test environment
(oxidation of graphite susceptor)
 Samples at temperature for short period of time (~20 min) which may not be long
enough for oxidative release
CEA-CENG Experiments

HEVA program (1983-1989)
 Heated Zircaloy-clad specimens of irradiated PWR fuel in mixtures of steam/H2
and pure H2 from 1800-2370 K (8 tests)
 Gamma spectrometry measured FPR from fuel and transport
 Aerosols collected in heated (temperature varied) cascade impactor and filters
 Control rod materials used in HEVA-07 (Ag-In-Cd exclusively) and HEVA-08
(control rod and fuel material)
CEA-CENG Experiments Cont’d

VERCORS program (1989 to 1994)
 Spent fuel samples heated to maximum temperature of 2620 K (6 tests)


3 PWR pellets with 2 half pellets (depleted U at ends) in unsealed Zrly clad
Re-irradiated in SILOE reactor to restore short-lived FPs (I, Te, Mo, Ba, La)
 Post-test gamma scanning (including gamma tomography) (complete FP
mass balance)
 Results of FP behaviour:







Nearly complete release of volatiles (Cs, I, Te and Sb)
Te and Sb initially trapped in unoxidized cladding
Semi-volatiles (Mo, Rh and Ba) (~1/2 that of volatile release depending on atmospheric conditions)
 Increased Mo release in oxiding conditions  92% release (VERCORS 5) vs 47% (VERCORS 4)
 Increased Ba and Rh release in reducing conditions  45 and 80% of Rh and Ba, (VERCORS 4) vs
20 and 55% (VERCORS 5)
Low-volatile FPs and actinides between 3 to 10% (Ru, Ce, Np, Sr and Eu)
Increased Np and Ce release under reducing conditions (VERCORS 4)
No release of non-volatile FPs (Zr, Nb La and Nd)
No significant enhancement in release in VERCORS 6 (early fuel collapse and partial liquid corium)
 Similar problems due to flow bypass (as for ORNL tests)
CEA-CENG Experiments Cont’d

VERCORS HT and RT program (1996 to 2002)
 Study FP and actinide release during later phase of accident with fuel
liquefaction
 Study FPR behaviour as influenced by:





Fuel type (UO2 versus MOX)
Fuel morphology (intact pellets versus debris fragments)
Fuel burnup
Presence of control materials (Ag, In, Cd and boric acid)
Environmental conditions (oxidizing or reducing conditions)
 Nb and La release in severe VERCORS HT and RT tests
 Fuel collapse temperature (2400 to 2600 K for fuel burnups of 47-70
GWd/tU) ~500 K below UO2 melting temperature
AECL-CRL Experiments

CRL Program (> 300 annealing tests)
 FPR from clad & unclad spent fuel samples (800 to 2350 K in Ar/H2,
steam and air atmospheres)
 Bare UO2 fragments (0.2-1.5 g each) and cladded specimens (Zircaloy foil
bags and short segments of Zircaloy-clad fuel with end caps)
–
Presence of Zircaloy can inhibit/delay release of volatile FPs
– Associated with time to oxidize Zircaloy cladding.
– Volatile FP release rates almost independent of temperature from 1670 to 2140 K after complete
clad oxidation
 FPR Behaviour:




Deposition and transport of FPs studied.
Volatiles release (Kr, Xe, I, Cs and Te) low in inert/reducing atmospheres but increase significantly after
clad oxidation in oxidizing atmospheres
Large fuel volatilization in high temperature tests with bare fuel (highly oxidizing environments)  lowvolatile release (Zr, La, Ba, Ce, Pr, Eu) via “matrix stripping” process
Eu, Ba released in hydrogen-rich atmospheres vs Mo, Ru, Nb in steam
 Oxygen potential of environment well characterized


Models developed for steam and air oxidation of UO2
Significant release of fission products in air (Ru, Nb)
Degraded Core Accident Phenomena
Temperature (K)
3120
2960
2870
2670
2245
2030
1720
1573
1470
1400
1220
(a)
1073
Degraded Core Accident Phenomena:
Pressure
 Comparable behavior
 Phebus FP (~0.2 MPa), LOFT FP-2 (~1 MPa), CORA (~0.2 to 1
MPa), TMI-2 (5 to 15 MPa)
 Enhanced clad ballooning & failure (low pressure)
 FPT-0 (trace-irradiated) at 735C
 Gap release measured (SFD, LOFT FP-2, Phebus FP)
 Aerosol composition
 Phebus FPT-0 and FPT-1: control rod (Ag,In,Cd), thermocouple
(Re), fuel rod (Sn,U) materials (~0.2 MPa)
 PBF SFD 1-4: FPs more important role (~7 MPa)
Degraded Core Accident Phenomena:
Control Rod Effects
 Pressure-Dependent Phenomena:
 (i) Low Pressure:
 SS clad/Zry guide tube contact with ballooning (high Cd vapour pressure) 
liquid phase ~1150C
 Ejection of molten control rod material  chemically dissolves guide tube/
clad of surrounding rods well below Zry MP (1760C) (CORA tests)
 (ii) High Pressure:
 Failure of SS clad at MP (~1450C)
 Phebus FP tests consistent with low pressure scenario

Control rod failure at 1120°C (FPT-0) and 1350°C (FPT-1)
Degraded Core Accident Phenomena:
Metallic Melt Formation
 Interaction of spacer grids/Zry cladding/control materials
 Relocation
 PBF SFD & TMI-2 accident (below coolant level)
 LOFT FP2, Phebus FP & CORA (cooler bundle regions)
 Freezing temperature of melt:
 1070 K (Ag-In-Cd alloy) to 1220 K (Zr-Fe eutectic), 1230 K (Zr-Ni
eutectic & elemental silver) and 1460 K (Zr-Ag eutectic)
 Metallic blockages in integral tests similar to TMI-2 but not as
extensive (shorter duration)
 FPT-0 metallographic examination
 Demonstrates role control rod plays in early melt formation
 Attack of Zry clad by molten Ag-In-Cd alloy
 Zr (20-40 wt%), Ag (10-50wt%), In (10-40wt%), U (<15wt%), O
(<10wt%), SS (<5wt%)
Degraded Core Accident Phenomena:
Zircaloy Oxidation
 Exothermic reaction (6.5 kJ/g-Zr oxidized)
 Accelerated heatup rates (10 K/s) at T > 1500 to 1700 K
 In-pile tests (PBF SFD, LOFT FP-2 and Phebus FP) and out-ofpile tests (CORA)
Degraded Core Accident Phenomena:
Hydrogen Generation

Related to steam availability
Test
Oxidation
of Zry (%)
Time
(s)
LOFT FP-2
49
~300
PBF-ST
Phebus FPT-0
75
85
~600
b
~1200
a. T> 1700 K; b. During oxidation phase.
a
a
Hydrogen Partitioning
Before
After
reflood (%) reflood (%)
20
80
77
b
82
23
-
Degraded Core Accident Phenomena:
Fuel Liquefaction
 UO2 & ZrO2 dissolved by metallic Zircaloy/-Zr(O)
(~1760 to 2000°C)
 Fuel liquefaction in integral tests
 PBF SFD tests (15-18%), LOFT FP-2 (15%)
 TMI-2 core (~45%)
 Phebus FPT-1 (20%), FPT-0 (50%)
 Separation between ceramic and metallic blockage
 (U,Zr)O2 freezes at higher temperature
Degraded Core Accident Phenomena:
Molten Pool Formation
 Ceramic heatup
 Steam diversion around blockage & poor thermal conductivity
 FP decay heat (TMI-2) or increased fission/electrical heating (integral tests)
 Molten pool (surrounded by U-rich crust) under a cavity
 TMI-2: (U,Zr)O2 ceramic with transition metal oxides (Cr2O3, Fe3O4)
~2700 K
 Phebus FPT-0: (U,Zr)O2 lattice with U (62wt%), Zr (22 wt%) & O(14wt%),
Fe(0.6wt%) ~2720 K
 Fuel movement
 TMI-2: Thermo-mechanical failure of crust  20 t lower plenum
 FPT-0: Downward motion of pool (from lower grid spacer) (18100 s)
 Comparable to MP tests
Degraded Core Accident Phenomena:
Debris Bed Formation
 TMI-2
 Top of molten pool & lower plenum region
 Integral tests
 Upper debris bed formed by coolant injection
(fragmentation) (SFD-ST & LOFT FP-2)
 Less steam-rich transients - decladded fuel/fragments in
upper part of bundle due to melting/relocation of clad
(SFD 1-4, Phebus FP)
FPR Behavior
 Comparison:
 PBF SFD-ST (steam-rich)/SFD (steam-starved), Phebus FPT-1
(steam-rich), TMI-2






Ce & actinides (typically < 0.01%)
Ru,Sr,Sb (<1%)
Ba (few %)
Mo (up to 50%)
Te (between 1 to 83%)
I,Cs, Noble Gas (up to ~90%)
 Comparable to annealing tests (ORNL, CEA, CRL)
FPR Phenomena
 FP trapping
 Te release (clad-oxidation state): Sn segregation  SnTe release
 Sb sequestered in metallic melts (Ni, Ag alloys)
 Burnup
 Enhanced volatile FPR in SFD 1-4 (high-burnup) vs SFD 1-1 (traceirradiated)
 Swelling (irradiated rods) in Phebus FPT-1
 Oxygen potential (H2/H2O ratio)
 Low Ba, Sr (Eu) release in steam tests (low- volatile oxides/hydroxides)
vs higher release in reducing tests (ST, VI, HEVA, VERCORS, CRL)
 Enhanced Ba release in FPT-0 during escalation phase (H2 generation)
 Low Ru release (PO2 too low to form high-volatile oxides)
 Low actinide release (fuel volatilization (UO3) ~ pO )
2
FPR Phenomena Cont’d
 FP behaviour in molten pool
 No enhancement with fuel liquefaction (non-coherent process)
 Volatile FPs: bubble nucleation, coalescence/growth & release via
buoyancy
 Bubble trapping at pool/surrounding crust
 Volatile FPs (I,Cs) in previously molten ceramics in PBF SFD,
Phebus FP & TMI-2 reactor
 TMI-2: Iron oxides in melt (lower limit of -120 kJ/mol)
 La, Ce, Sr as oxide (soluble in (U,Zr)O2)
 Ru, Sb metal immiscible in ceramic melt
 Cooldown/reflood
 Small release (fuel relocation) in Phebus FPT-0 vs large release in
LOFT FP-2 (~12% volatile release) (local heating)
Ba Release in ORNL, CEA and Phebus
Tests
Test
Temp. (K)
Duration (min)
Atmosphere
Ba Release (%)
HI-4
HI-5
VI-2
VI-3
VI-4
VI-5
2200
2025
2300
2700
2440
2720
20
23
60
20
20
20
H2O
H2O
H2O
H2O
H2
H2
<1
<1
19
30
27
76
HEVA-4
HEVA-6
VERCORS-1
VERCORS-4
VERCORS-5
VERCORS HT-1
2270
2370
2130
2570
2570
3070
7
30
17
30
30
7
H2O+ H2
H2
H2O+ H2
H2
H2O
H2
6
27
4
80
55
49
Phebus FPT-0
PEBUS FPT-1
~2700
~2500
-
H2O/H2
H2O/H2
1
1
Difference in Ba Release from OutPile and Phebus Tests

Short duration of temperature escalation in in-pile tests
 No “high temperature plateau” as in annealing tests but rather
temperature escalation (with formation of molten pool) in Phebus
test

Ba volatility is reduced with significant amount of ZrO2 in
fuel melt (~47 mol%) and small amounts of iron oxide in
in-reactor test

Thermochemical analysis with GEMINI2
 Reduced Ba vapor pressure in solidus-liquidus transition zone in
the U-Ba-O phase diagram (~2400-3100 K)
Concluding Remarks
 In- and out-pile experiments on severe accident melt progression &
FPR behaviour reviewed
• Melt progression non-coherent process
• Phebus FPT-0 and -1 tests performed for longer high-temperature period
than earlier in-pile experiments  information on late-phase behaviour
• Local propagation of core melt due to control rod failure (at lower
temperature)
• Metallic blockages result from interactions of spacer grids, fuel rod
cladding material and control rod materials that flow down bundle and
solidify at lower (cooler) position
– Separation between metallic and ceramic blockages with freezing of (U,Zr)O2
melt at higher temperature
– Observed melting temperature of ceramic blockage in Phebus FPT-0 test
(~2720 K) slightly lower than pure ceramic (~2800 K) due to eutectic
interaction (consistent with TMI-2 examination)
– Molten pool formed due to increased fission heat generation and reduced heat
transfer in several in-pile experiments (held in place by ceramic crust)
Concluding Remarks Cont’d
 Consistent release behaviour of volatile (Xe, Kr, I, Cs, Te and Sb),
semi-volatile (Mo, Rh, Ba), low volatile (Ru, Ce, Np, Sr and Eu) and
non-volatile (Zr, Nb, La and Ne) FPs observed in annealing (ORNL,
CEA, CRL) and in-pile tests
 Reduced volatility of Ba for in-reactor experiments (thermochemical
effects with presence of iron and zirconium oxides)
 Local atmospheric condition/oxygen potential influences low-volatile
fission product release behaviour
 Non-coherent melt progression masks individual release mechanisms
identified in out-of-pile experiments
 Enhanced release due to fuel liquefaction not typically observed in
separate effects experiments
Acknowledgements


Discussions with P. Elder, L. Dickson, M.
Schwarz, R. Zeyen, B. Clement, M. Kissane and
G. Ducros
Funded by NSERC/COG Collaborative Research
and Development (CRD) Grant