performance assessment of mox fuel with 20% cold … assessment of mox fuel with 20% cold-worked...
Post on 21-Mar-2018
216 Views
Preview:
TRANSCRIPT
Performance assessment of MOX fuel with
20% cold-worked alloy D9 cladding and
wrapper irradiated in FBTR
Jojo Joseph, Divakar Ramachandran, C. N. Venkiteswaran, V. Karthik, T. Johny, B. P. C. Rao, T. Jayakumar
Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603102, India
E-mail: divakar@igcar.gov.in
International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios, 4 – 7 March, 2013, Paris, France
Introduction
Fast Breeder Test Reactor (FBTR) at Kalpakkam, India is used as an irradiation facility
Prototype Fast Breeder Reactor (PFBR) is in an advanced stage of construction at Kalpakkam
PFBR is designed to operate with
Mixed-oxide (MOX) annular fuel pellets
20% cold-worked Ti-modified austenitic stainless steel (alloy D9) cladding and wrapper
Target burn-up of 100 GWd/t
Objective: Performance evaluation of fuel for a burn-up exceeding 100 GWd/t
One fuel sub-assembly with 37 short-length MOX test fuel pins was irradiated in FBTR
Number of fuel pins 37
Fuel 29% PuO2, Rest UO2 with 53.5% U233
O/M 1.98 to 2.00 Fuel stack length 240 mm Type of bond Helium Linear mass g/cm 2.1 – 2.2 Fuel Density (% TD) 91 ± 1 % Pellet /Central hole dia. 5.52 mm / 1.75 mm Outer diameter of fuel pin 6.6 mm Inner diameter of fuel pin 5.7 mm Clad & Wrapper material 20 % CW Alloy D9
PFBR MOX Test Fuel Sub-assembly
Irradiation History
Total duration ~ 7.5 years
Peak Burn-up 112 GWd/t (10.3 atom percent)
Peak linear heat rate
450 W/cm
Peak power of reactor
18 MWt
Peak displacement
damage 62 dpa
Sodium Inlet/Outlet
temperatures
350 °C / 435 °C – up to 81 GWd/t
380 °C / 480 °C – until discharge at
112 GWd/t
Location of FSA in FBTR core
PIE Facility at IGCAR
The hot-cells of RML are α/β/γ leak-tight, concrete shielded, inert atmosphere cells equipped with a range of PIE equipment, many of which are designed in-house
The FSA was received into the hot-cells and cleaned for removal of sodium using high purity ethanol. Visual examination and remote metrological measurements were carried out before dismantling and extracting the fuel bundle
Post-Irradiation Examination
The fuel pins were subjected to non-destructive evaluations by profilometry, eddy current testing, X-radiography, neutron radiography and gamma scanning
Selected fuel pins were subjected to puncture test for fission gas analysis and sectioning for metallographic examinations respectively
The specimens were subjected to swelling measurements by liquid immersion method within the hot-cell
Clad tube specimens were subjected to tensile test after removal of the fuel by dissolution
Flat tensile specimens extracted from the wrapper remotely using a CNC machine were subjected to tensile tests
Dimensional Measurements
Increase in the corner-to-corner distance maximum 0.31 ± 0.02 mm at the core centre (0.54%)
Effect of swelling alone
Increase in the width across flats maximum 0.4 mm in the core centre region (0.80%)
Effect of swelling and irradiation creep
Irradiation creep does not significantly contribute to the total strain for wrapper
He
ad-to
0fo
ot M
isalignm
en
t 1
.8 m
m
Location of subassembly in
the core as viewed from top
head endfoot end
head endfoot end
Bowing along orientation - I
Bowing along orientation - II
Quantification of head-to-foot
misalignment
Profilometry of fuel pins Peak swelling at mid wall T ~
500°C
Cladding swelling is higher due to fuel adjacency effect causing higher T and T gradient across cladding wall
Swelling component was only about 40%; irradiation creep dominates (possibly due to fission gas pressure) In
crea
se in
dia
met
er m
easu
red
by
pro
filo
met
ry 5
6 –
91
um
Tensile testing of clad and wrapper
Recovery of ductility at irradiation temperatures beyond 480°C (upper portions of fuel column) is seen in both clad and wrapper
The uniform elongation at operating temperature is lower for cladding (~ 3%) as compared to wrapper (~4.5%)
Trends are similar to those reported in the literature
Summary – Structural Materials
Both D9 cladding and wrapper show similar trends
For low irradiation temperatures, there is significant hardening with a decrease in uniform elongation; effect is more prominent in wrapper due to lower temperatures
Hardening effect decreases with increase in irradiation temperature
Swelling does not seem to have influenced the degradation of mechanical properties
Overall, for the Alloy D9 cladding and wrapper irradiated to 60 dpa in FBTR, swelling is low (2% for clad and 0.2% for wrapper) and
There is adequate retention of the mechanical properties of clad and wrapper at irradiation temperature
Dimensional changes, fission gas release, fission products and microstructural evolution
PIE OF IRRADIATED FUEL
Correlation between diverse NDE techniques
N-Radiograph
Gamma scan (Cs137)
Eddy current profile
200 µm
microscopy
Central hole diameter increase - densification (porosity migration) Fuel-clad gap – No FCMI
Central hole shrinkage - fuel swelling (Retention of FP and absence of clad swelling) Fuel-clad gap – closed, FCMI
500 µm
Peak power location Core top location
Pellet-Pellet interface e
500 µm
Axial variation in central hole dia
Irradiated fuel cross-sections
Fission Gas Analysis
The fission gas release measurements on four fuel pins indicated internal pressure of 2.4 - 2.8 MPa at ambient temperature
The fission gas release is in the range of 82-85% typical of high burn-up mixed oxide fuels
High fission gas release of about 85% and a plenum pressure of 6 – 7 MPa at operating temperature support the earlier finding that about 60% of the pin diametral strain is attributed to irradiation creep.
Reference volume
Sample vial
Fuel pin
Puncture tool
Summary
Cladding
2% swelling strain with major contribution due to irradiation creep
Hardening with residual ductility >3% at 60 dpa at high-temperatures
Wrapper
0.2% swelling strain with major contribution from void swelling
Hardening with residual ducility ~ 5% for peak dpa
MOX Fuel
Low swelling
Over 85% fission gas release leading to plenum gas pressure ~ 2.8MPa
migration and deposition of fission products
Microstructure variations as per temperature, temperature gradients in fuel
Conclusions
The low swelling of fuel, high fission gas release and fuel microstructural evolution indicate safe and satisfactory performance of the MOX fuel up to the 112 GWd/t achieved in the present campaign
The Alloy D9 cladding and wrapper have performed satisfactorily with respect to swelling resistance and retention of mechanical properties at a displacement damage of 60 dpa achieved in the present study
The PIE results validate the fuel and structural material design as well as the fabrication and quality control (QC) routes adopted for PFBR
FCCI observed would need to be considered for enhancement of burn-up
top related