Small Punch Test (SPT) is one of the miniature test techniques used for evaluating tensile properties when limited dimension of material is available for standard mechanical test or handling of large specimen is difficult, especially while dealing with irradiated materials. In order to use SPT for determination of mechanical properties, it is crucial to establish correlation coefficients by comparing results obtained by SPT with that from standard tensile tests. Present work involves developing correlation coefficients and evaluation of tensile properties of reactor pressure vessel (RPV) steel using SPT technique. Specimen dimensions used for SPT was 10x10x0.5 mm3. These specimens were prepared from Charpy V-Notch specimens, which were earlier subjected to standard impact testing. Correlation coefficient for RPV material was established by comparison of SPT data with the tensile properties obtained using standard tensile test. Subsequently, SPT data was acquired on RPV thermal material and tensile properties were obtained using the correlation coefficient for validation. This paper describes the SPT set-up used during the current studies and methodology used for establishing the correlation coefficient on RPV base material.

Carbon steel straight pipes form part of the primary heat transport system of nuclear power plants. They may be subjected to dynamic-cyclic loads during earthquake event. This dynamic cyclic loading significantly affects the fracture resistance and the load carrying capacity of cracked components. Further, for the application of leak-before-break concepts, experimental investigations are essential. Therefore, the aim of the present study is to investigate the effect of cyclic loading on fracture resistance of through-wall cracked straight pipes. The paper presents details of the cyclic fracture test on a straight pipe and the results thereof. The behaviour of the pipe was compared with the fracture behaviour under monotonic loading. The study has shown that the load capacity of the pipe is reduced under cyclic loading and the failure is by cyclic tearing followed by sudden collapse.

Nodular corrosion can severely affect the service life of zirconium alloys in water cooled nuclear reactors. In this study, the susceptibility of pressure tube alloy to undergo nodular corrosion in water-steam with air surroundings at 250°C and 300°C is investigated. Accelerated corrosion has been studied in single phase steam at 500°C. The effect of specimen surface condition and cold work on nodular corrosion susceptibility has been studied in detail. The results indicate that surface condition of the specimen plays primary role in determining the propensity of the alloy to undergo nodular corrosion. Cold work has negligible effect on the initiation as well as growth of oxide nodules in Zr-2.5 (wt. %)Nb alloy. The effect of temperature is highly localized and is visible on the size of the nodules. Presence of non-uniform or discontinuous autoclave layer increases the susceptibility of pressure tube alloy to undergo nodular corrosion.

The level-3 Leak-Before-Break (LBB) assessment of nuclear piping system is based on simplified fracture mechanics calculations. A postulated through-wall circumferential crack is shown to be stable even under extreme loading conditions. An important aspect of this fracture assessment, which is generally not considered, is the reduction in load at the cracked section that occurs due to the local and global compliance of the piping system. In this study, the effect of piping system compliance, crack size and crack location on the load carrying capacity of SA-333 Gr. 6 carbon steel pipes was analysed. Experimental investigations were performed on 8 NB Schedule 100 pipes (Outer diameter 219 mm and wall thickness of approx. 12.6 mm) under quasi-static four point bending loads. Out of the six fracture tests, one specimen had a through-wall circumferential crack at the centre (along the length of pipe) and the remaining five tests had off-centred cracks. While the initial crack size showed a weak influence on the load carrying capacity of a stiff piping system, the effect of crack location was, however, quite significant. Calculations based on twice-elastic slope (TES) method and limit analysis of uncracked piping system (in some cases) for evaluation of load carrying capacity of pipes under finite compliance may lead to large safety margins.

Uranium rich U-Zr-Mo alloys are considered to be good alternative fuel for fast reactors due to their many favourable properties to achieve higher burnup. Studies related to structural and microstructural characterization, phase transformation, thermal expansion and microhardness were carried out on U-6Zr, U-4Zr-2Mo, U-2Zr-4Mo and U-6Mo alloys (composition in wt%) where the total amount of alloying elements was restricted to 6 wt%. U-6Zr alloy is considered here as the reference composition and the effect of Mo addition in the U-Zr alloy has been investigated. The phase stability of alloys after water quenching and furnace cooling from high temperature bcc gamma-phase was studied. The alloys undergo various transformations from high temperature bcc gamma-phase to a variety of equilibrium and intermediate phases depending upon alloy composition and cooling condition. A significant change in the crystal structure, microstructure, phase transformation, solidus temperature, thermal expansion and hardness data was observed when alloying element Zr is gradually replaced with Mo. (C) 2020 Elsevier B.V. All rights reserved.

Uranium-rich U-Zr-Nb and U-Zr-Mo alloys are considered to be good alternative fuel for fast reactors. From that prospective, number of selected uranium rich binary and ternary U-6X (X= Zr + Nb/Mo) alloy samples (i.e., U-6Zr, U-6Nb, U-2Zr-4Nb, U-4Zr-2Mo, U-2Zr-4Mo) with total alloying content of 6 wt% were prepared and characterized by X-ray diffraction (XRD), scanning electron microscope (SEM) equipped with energy dispersive spectroscope (EDS) and differential thermal analyzer (DTA). Molar heat capacity is considered as one of the most important thermophysical properties of nuclear fuel. The molar heat capacity measurements of U-6Zr, U-2Zr-4Nb, U-6Nb, U-4Zr-2Mo and U-2Zr-4Mo alloys were carried out using differential scanning calorimeter (DSC) in the temperature range of 323-823 K. The U-6Zr alloy is considered here as the reference composition and the effect of Nb or Mo addition on the molar heat capacity data of U-6X alloys has been highlighted. The thermodynamic functions such as molar enthalpy increment H-m(0) (T) - H-m(0)(298.15K) and molar entropy increment S-m(0)(T) - S-m(0)(298.15K) of the alloy samples have been evaluated using measured molar heat capacity C-p,m(0)(T) data. (C) 2020 Elsevier B.V. All rights reserved.

Fracture studies were carried out on narrow gap welded SA 312 Type 304LN stainless steel straight pipes having circumferential through-wall notch in the weld. Totally six specimens were tested; one specimen was subjected to monotonic loading and the remaining five specimens were subjected to quasi-cyclic loading. Under quasi-cyclic loading, one specimen was tested under incremental displacement-controlled loading and the remaining four specimens were tested constant amplitude load-controlled loading. Based on the load-controlled experimental results, a plot of the cyclic load amplitude versus number of cycles to failure of the specimens was obtained. The experimental results indicate that the piping components subjected to quasi-cyclic loading may fail in very less number of loading cycles even when the load amplitude is sufficiently below the monotonic fracture/collapse load. Crack growth behaviour and cyclic tearing instability in the specimens, which are required for realistic assessment of Leak Before Break applicability, was studied.

In case of severe accidents like loss of coolant accident with failure of emergency core cooling system, the pressure tubes get heated quickly causing severe deformation. In this paper, an analytical solution technique has been used for calculating the radial deformation behaviour of pressure tubes. The analysis is carried out for different cases when the pressure tube is subjected to different values of internal pressure and heating rates. In the absence of specimen level creep test data of Indian pressure tube, data of Canadian pressure tube of similar material has been used. Due to observed differences between experimental and analytical values of radial deformation with time, the actual parameters have been calibrated from component level experiments using an optimization algorithm. Both uniaxial and bi-axial states of stress have been considered. Finite element method has also been used to calculate radial deformation of pressure tube and compared with experiment.

Frieswal cattle is one of the crossbred strains having 5/8 Holstein Friesian and 3/8 Sahiwal inheritance, develo pedby ICAR-Central Institute for Research on Cattle, Meerut, in collaboration with Ministry of Defence. The projectwas started in 1987 for the evolution of a new crossbred national milch breed: Frieswal – (Holstein×Sahiwal), yielding 4,000 kg of milk with 4% butter fat in a mature lactation of 300 days. It is the only project in the country where such a huge population of animals (more than 20,000) is available with accurate production and pedigree records of each and every animal over generations, with well-established progeny testing network in 37 military farms to cover all the agro-climatic zones of the country. In this review, an attempt has been made to analyze the genesis, present status and future prospects of Frieswal cattle in India.

The chemical interaction of U-Zr based fast reactor fuel with T91 grade steel cladding was investigated by employing DTA (differential thermal analysis), diffusion couple experiment and SEM/EDS, SEM/EBSD analyses. The DTA results revealed a strong exothermic reaction just below the solidus temperature of the U-Zr alloy. The eutectic liquefaction temperature between U-Zr alloy and T91 steel was found to be at 995 K. The interdiffusion between U-10Zr and T91 resulted in the formation of UFe2-type phase, a Zr-rich layer and a Zr-depleted layer at the interface. The growth kinetics of the reaction zone was also determined using the total width of UFe2 and Zr-rich reaction layers. The activation energy for the reaction was found to be 170 kJ mol(-1). The Zr-rich layer at the interface acts as diffusion barrier which enhances the fuel-cladding chemical compatibility at reactor operating temperature. However, the U-Zr fuel is no more compatible with the cladding above the eutectic liquefaction temperature (995 K). (C) 2018 Elsevier B.V. All rights reserved.

The All India Co-ordinated Research Project (AICRP) on Cattle is implemented in the home tract of Gir cattle for their conservation, propagation and genetic improvement. The involvement of Gir animal owners under the programme has helped to create awareness among the farmers on the importance of this valuable breed. The production of frozen semen doses of genetically superior young bulls and the availability of AI facility at the doorsteps of the farmer has increased the conception rate to a large extent. All lactation production performance of Gir animals maintained in the GP unit revealed better performance than the first lactation production traits. The overall total lactation milk yield, 305-days yield, lactation length, peak yield and dry period were estimated as 3801.0 kg, 2868.30 kg, 457.2 days, 14.80 kg and 68.50 days, respectively. The mean all lactation service period and calving interval were 216.50 and 498.50 days, respectively. The production and reproduction performance of Gir cattle has improved over the years due to the implementation of the project and continuous efforts will be made to propagate this valuable germplasm to the needy people of the country.

This paper presents outcome of round robin exercise on usage of ball indentation technique to predict strength properties of three different materials which are used in Indian nuclear reactors. The main objectives were to evolve a standardized procedure of determining the mechanical properties by the Ball Indentation technique and quantification of the variation in strength and fracture properties due to ageing through such tests. Standard uniaxial tensile tests were also carried out to quantify the differences between strength properties predicted by ball indentation and those determined using conventional tests. Three different materials studied were, carbon-manganese steel (Grade: SA 333Gr.6), stainless steel (SA312 Type 304LN) and zirconium alloy (Zr-2.5Nb). The carbon-manganese steel and stainless steel materials were drawn from extruded pipes while Zr-2.5Nb was drawn from pressure tubes, used in Indian Pressurized Heavy Water Reactors. Ball indentation tests were carried out on materials in as received and in aged conditions. The ageing was simulated artificially. The carbon-manganese steel and stainless steels were subjected to cold work while Zr-2.5Nb alloy was subjected to heat treatment at different temperatures with varying hold times and in some cases with charged hydrogen. The round robin exercise was conducted within different research centers of Department of Atomic Energy (DAE), India. This was a blind exercise in which all the conventional tests were carried out by a group (who was not involved in Ball Indentation tests) after the completion of ball indentation tests by participants. The paper presents the prediction of strength properties, using ball indentation, by different participants. The reasons of scatter in the results among the various participants and differences with respect to conventional test results are discussed in the paper.

Full field blast experiments were performed to assess the potential of fluid filled polymer foam for blast mitigation. The experiments involve air blast loading of clamped mild steel plates covered with fluid filed polymer foam for blast protection. The deformation profiles and maximum deflections of plates are compared with and without foam protection. The experimental results indicate a reduction in the plate deflection up to 50% with foam protection.

The aim of this study was to study the effect of loading frequency on the fatigue crack growth rate (FCGR) behaviour of Type 304L(N) austenitic stainless steel material. Compact tension (C(T)) specimens machined from the solution annealed pipe were used for carrying out the FCGR tests as per the standard procedure of ASTM E647. Constant amplitude loading method with varying frequency (0.01–10 Hz) was followed for cyclic loading during the tests. The tests were carried out at load ratio 0.1 and ambient temperature. Experimental results show that FCGR increases with decrease in frequency for a given stress intensity factor range. Increase in FCGR is very significant from high to low frequency. Fracture surface examinations of tested specimens were carried out using scanning electron microscope to understand the increased crack growth at lower frequency. The fracture surfaces revealed the presence of facets which have feature of brittle fracture. These facets could be the locations for the formation of deformation-induced-martensite (DIM). The facets are relatively harder (320 HV) as compared to matrix of the material (155 HV) which may have led to higher crack growth rate. Area covered by these facets is more for fracture surface tested at lower frequency. FCGR of present test study is also compared with FCGR curve for austenitic stainless steel given in ASME Section XI. ASME FCGR curve is comparable with that of the FCGR curve of loading frequency 10 Hz. However, ASME FCGR would lead to over-prediction of fatigue crack growth life for lower loading frequency scenario.

This paper presents the results of the investigation on the deformation and rupture characteristics of Indian pressurized heavy water reactor (IPHWR) fuel pins under simulated loss of coolant accident (LOCA) condition in steam environment. Transient heating experiments were carried out on single fuel pin internally pressurized with argon gas in the range 3-70 bar. Effect of internal pressure on burst temperature, influence of burst temperature on the circumferential strain and rupture opening area were also studied. Two circumferential strain maxima at the burst temperatures of 740 & similar to 979 degrees C and a minimum at the burst temperature of similar to 868 degrees C were observed. It was found that oxidation had considerable effect on the burst behavior. Test data were used to derive a direct empirical correlation for burst stress exclusively as a function of temperature. The ballooning and rupture behaviours in steam and argon environments have been compared. Experimental data were examined against various correlations using Erbacher equation and author's previous correlation in argon. A second burst correlation has also been developed combining the equation in argon from the previous work of the authors and an exponential factor with oxygen content as a parameter assuming the burst stress to be a function of both temperature and oxygen concentration. The burst temperatures predicted by this empirical correlation are in good agreement with the test data. (C) 2017 Elsevier B.V. All rights reserved.

The present study involves the estimation of ring tensile properties of Indian Pressurised Heavy Water Reactor (IPHWR) fuel cladding made of Zircaloy-4, subjected to experiments under a simulated loss-of-coolant- accident (LOCA)condition. Isothermal steam oxidation experiments were conducted on clad tube specimens at temperatures ranging from 900 to 1200 degrees C at an interval of 50 degrees C for different soaking periods with subsequent quenching in water at ambient temperature. The specimens, which survived quenching, were then subjected to ambient temperature ring tension test (RTT). The microstructure was correlated with the mechanical properties. The yield strength (YS) and ultimate tensile strength (UTS) increased initially with rise in oxidation temperature and time duration but then decreased with further increase in oxidation. Ductility is adversely affected with rising oxidation temperature and longer holding time. A higher fraction of load bearing phase and lower oxygen content in it ensures higher residual ductility. Cladding shows almost zero ductility behavior in RIT when load bearing phase fraction is less than 0.72 and its average oxygen concentration is greater than 0.58 wt%. (C) 2017 Elsevier B.V. All rights reserved.

In the present work, a coupled finite element (FE) non-linear grided and element free Galerkin (EFG) technique has been projected to evaluate the fatigue life of cracked heterogeneous plate. The discontinuities positioned away from the major crack have insignificant effect on the fatigue life. So, the small nearby region is modeled by EFGM as a non-homogenous medium, rest of domain using conventional FEM except interface elements applied in the transition zone. This will exploit the advantage of both the methods. In this method, domain is discretized by non consistence mesh, fine mesh used for multiple discontinuities and coarse mesh with equivalent properties distribution. Strain energy approach is used to calculate equivalent properties of the region of interest. Displacement compatibility of the intersection nodes between meshing difference is achieved by special transition elements. This non-linear grid approach is six times more efficient as compared to linear grid approach. The results obtained by proposed method are found in good agreement with those obtained by analytical/reference solutions.