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 changes and Mechanical Properties

PIE OF STRUCTURAL

MATERIALS

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

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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

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Comparison of Swelling with SS316

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

Trends in tensile properties

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

Non-destructive examination of fuel pins R

adio

grap

hy

ECT

Gam

ma

X-ray

Neutron

Cs137

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

Fuel microstructure and interaction with clad

Wall thickness reduction

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

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