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Fuel Clad Chemical Interaction (FCCI) is one of the life limiting issues in the MOX fuel pins of fast breeder reactors. Clad wastage due to FCCI coupled with stress arising from fission gas pressure and Fuel Clad Mechanical Interaction (FCMI) due to fuel swelling can lead to fuel pin failure. Non-destructive evaluation (NDE) techniques such as Gamma Scanning, Eddy current testing and Neutron Radiography have been successfully used on MOX fuel pins of a test sub-assembly irradiated in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GW d/t to detect signatures of FCCI. The results obtained by the various NDE techniques have been correlated and verified through metallographic examination on fuel pin cross-sections. The measured clad wastage of 85 μm agrees well with a model developed for MOX fuel with high Pu content. The results of examinations have enabled validation of the model and given confidence to the designer that PFBR MOX fuel can safely attain the target burn-up of 100 GW d/t.

Fast reactor cores employ a large number of sub-assemblies for neutron and gamma shielding constituting more than 60% of the core volume. Use of effective and economical shielding materials can significantly improve the safety and cost competitiveness of fast reactors. Ferroboron has been identified as an efficient neutron shielding material for deployment in future commercial breeder reactors in India that can lead to significant cost savings. After successful shielding experiments and out-of-pile compatibility tests, a SS 304L clad irradiation capsule containing ferroboron powder has been subjected to an accelerated irradiation test in Fast Breeder Test Reactor (FBTR) to a peak neutron fluence of 8.18E21 n/cm2 for evaluating its in-reactor performance. Results of Post Irradiation Examinations carried out on the irradiation capsule to assess the various performance parameters such as slumping of ferroboron stack, helium release and chemical interaction of SS 304L with ferroboron are reported here. The examinations have indicated that extent of slumping is less than 1% and pressure build up due to helium release is 0.17 MPa (at room temperature), which are well within the permissible limits. A maximum clad wastage of 20% of wall thickness has been observed due to chemical interaction with lithium produced from (n,α) reaction. The irradiation performance assessment of ferroboron indicates its suitability for in-core shielding application in fast reactors.

The dimensional measurements such as fuel stack length, fuel pin diameter, pellet-to-pellet gap, plenum and spring length of 112 GWd/t irradiated Uranium Plutonium mixed oxide (MOX) fuel pins have been carried out using X-ray radiography. The X-ray radiography procedure is optimized for the exposure time, applied voltage, applied current to minimize the effect of high gamma radiations emitted from the irradiated fuel pins on the X-ray films. The dimensional measurements from the radiography images indicate a maximum increase in the stack length of 1.3%. The relatively low fuel swelling, compared to the carbide fuel, is attributed to the low fuel smear density and high fission gas release. The analyses of digitized radiography images revealed a few radial cracks in the fuel pellets. At the bottom portion of the fuel pins, pellet-to-pellet gaps and pellet-to-clad gaps are seen clearly. Our results indicate that after irradiation to 112 GWd/t, the spring length has compressed to ∼88.7 mm against the original value of 90 mm due to volumetric swelling of the fuel column. It is observed that the diametrical strain is highest at the central location of fuel pins and the spring column is compressed. The contrast sensitivity of the radiography images is enhanced by adopting image processing approaches. A 5 × 5 Laplacian kernel based edge detection technique is used to evaluate the fuel pin diameter accurately. The average increase in pin diameter, after irradiation, is found to be 50 μm.

Cold worked austenitic stainless steel type AISI 316 is used as the material for fuel cladding and wrapper of the Fast Breeder Test Reactor (FBTR), India. The evaluation of mechanical properties of these core structurals is very essential to assess its integrity and ensure safe and productive operation of FBTR to very high burn-ups. The changes in the mechanical properties of these core structurals are associated with microstructural changes caused by high fluence neutron irradiation and temperatures of 673–823 K. Remote tensile testing has been used for evaluating the tensile properties of irradiated clad tubes and shear punch test using small disk specimens for evaluating the properties of irradiated hexagonal wrapper. This paper will highlight the methods employed for evaluating the mechanical properties of the irradiated cladding and wrapper and discuss the trends in properties as a function of dpa (displacement per atom) and irradiation temperature.

Austenitic stainless steel SS316 and its variants are the common materials for the fast reactor structural components. Using the Fast Breeder Test Reactor (FBTR) as an irradiation test bed, a systematic analysis of the irradiation performance of the austenitic stainless steel has been undertaken. The performance of 20% cold worked SS316 has been assessed by examining the cladding and wrapper of FBTR at various displacement damages. The modified version of SS316, alloy D9, chosen for PFBR has been subjected to test irradiations in FBTR. Further modification of alloy D9 with respect to minor elements is also being studied The salient features of (i) mechanical and microstructural behaviour of SS316 at different fluence levels, (ii) the ongoing irradiation experiments on alloy D9 and (iii)microstructural studies on modified versions of alloy D9 are presented in this paper.

Tin Sulfide (SnS) nanosheets were synthesized by wet chemical route using ethylene glycol (EG) and without using any surfactant. Structural and phase purity were confirmed by powder X-ray diffraction pattern which shows the orthorhombic structure of SnS. The sheets like morphology and particle size of the synthesized product were identified by using analytical transmission electron microscope (TEM) and high resolution transmission electron microscope (HRTEM). Agglomeration of SnS nanoparticles was found to lead to the formation of nanosheets. UV-VIS-NIR absorption spectrum of SnS nanosheets shows the direct transition at 1.88 eV. Compared to bulk band gap a blue shift of 0.58 eV has been observed for direct transition. This is due to the quantum confinement effect. Room temperature photoluminescence spectrum of SnS nanosheets shows two emission bands at 1.75 and 1.57 eV respectively which are assigned to band gap and defect level transitions.

Fission gas extraction and analysis has been carried out on fuel pins of fast breeder test reactor subassemblies at different burn-ups as a part of post-irradiation examinations to characterise the performance of the unique mixed carbide fuel and estimate the residual life. This paper describes in detail the fission gas extraction and analysis carried out on fuel pins after various stages of burn-ups of 25, 50 and 100 GWd/t. The analysis of fission gas was carried out by gas chromatography. Analysis indicated that the maximum percentage of gas release is 14% after a burn-up of 100 GWd/t, and the corresponding plenum pressure is 1.2 MPa.

Fast breeder test reactor (FBTR) uses unique, indigenously developed mixed carbide of plutonium and uranium (70% PuC + 30% UC) as fuel. The performance of the fuel was studied at different burn-ups by carrying out post-irradiation examinations (PIE) to facilitate prediction and extension of life for the indigenously developed fuel. As a part of PIE, remote metallography was carried out to study the swelling behaviour of the fuel, estimate the fuel-clad gap and examine the fuel and clad microstructures. Results of the metallographic examinations were used to increase the limit of FBTR fuel burn up and linear heat ratings. This paper describes in detail the remote metallographic techniques adopted during PIE of fuel after a burn-up of 25 GWd/t and 50 GWd/t.

Post-irradiation examination (PIE) is a vital link in the nuclear fuel cycle for providing valuable feedback on the performance and residual life of the fuel and structural materials to designers, fabricators, and reactor operating personnel. The challenging task of setting up of α,β,γ inert atmosphere hot cell facility for PIE of Fast Breeder Test Reactor (FBTR) was accomplished successfully and irradiation performance of the FBTR mixed carbide fuel was assessed stage wise at various burnups starting from 25 GWd/t upto 155 GWd/t. With FBTR being used as a test bed for irradiation experiments on various FBR fuels and structural materials, PIE of various materials subjected to experimental irradiation like the PFBR MOX fuel, FBTR grid plate material have also been carried out to provide valuable feedback to the designers. This paper highlights the (i) results of comprehensive PIE carried on mixed carbide fuel & structural material (ii) control rod performance and (iii) outcome of the examinations on the experimental irradiated sub assemblies..