@article{morrison_gould_charit_hassan_2019, title={Performance Evaluation of Surface-Activated Solid-State Welding for ASTM A992 Structural Steel}, volume={31}, ISBN={1943-5533}, DOI={10.1061/(ASCE)MT.1943-5533.0002805}, abstractNote={AbstractThis paper presents the results of a pilot study to evaluate a solid-state welding technology, called surface activated solid-state (SASS) welding, for joining structural steel members. SAS...}, number={8}, journal={JOURNAL OF MATERIALS IN CIVIL ENGINEERING}, author={Morrison, Machel L. and Gould, Jerry and Charit, Indrajit and Hassan, Tasnim}, year={2019} }
 @article{charit_murty_2008, title={Creep behavior of niobium-modified zirconium alloys}, volume={374}, ISSN={["0022-3115"]}, DOI={10.1016/j.jnucmat.2007.08.019}, abstractNote={Zirconium (Zr) alloys remain as the main cladding materials in most water reactors. Historically, a series of Zircaloys were developed, and two versions, Zircaloy-2 and -4, are still employed in many reactors. The recent trend is to use the Nb-modified zirconium alloys as the Nb addition improves cladding performance in various ways, most significant being superior long term corrosion resistance. Hence, new alloys with Nb additions have recently been developed, such as Zirlo™2 and M5™3. Although it is known that creep properties improve, there have been very few data available to precisely evaluate the creep characteristics of new commercial alloys. However, the creep behavior of many Nb-modified zirconium alloys has been studied in several occasions. In this study, we have collected the creep data of these Nb-modified alloys from the open literature as well as our own study over a wide range of stresses and temperatures. The data have been compared with those of conventional Zr and Zircaloys to determine the exact role Nb plays. It has been argued that Nb-modified zirconium alloys would behave as Class-A alloys (stress exponent of 3) with the Nb atoms forming solute atmospheres around dislocations and thus, impeding dislocation glide under suitable conditions. On the other hand, zirconium and Zircaloys behave as Class-M alloys with a stress exponent of ⩾4, attesting to the dislocation climb-controlled deformation mode.}, number={3}, journal={JOURNAL OF NUCLEAR MATERIALS}, author={Charit, I. and Murty, K. L.}, year={2008}, month={Mar}, pages={354–363} }
 @article{gollapudi_charit_murty_2008, title={Creep mechanisms in TI-3Al-2.5V alloy tubing deformed under closed-end internal gas pressurization}, volume={56}, ISSN={["1873-2453"]}, DOI={10.1016/j.actamat.2008.01.052}, abstractNote={Creep tests were carried out on Ti–3Al–2.5V alloy tubing in the temperature range of 723–873 K under closed-end internal pressurization. The data thus obtained were analyzed to obtain the mechanistic creep parameters (stress exponent and activation energy). Transitions in creep mechanisms were noted as the stress exponent varied from a lower value of 1 through 2 to a higher value of 5 with increasing stress where the activation energy assumed values of 232 and 325 kJ mol−1, respectively. The creep mechanisms were elucidated in the light of standard creep models supported by the substructures studied by transmission electron microscopy. Newtonian viscous creep (n = 1) at lower stresses was identified to be in accordance with a slip band model named after Spingarn and Nix. Grain boundary sliding with n = 2 was noted in an intermediate stress region while climb of edge dislocations was observed to control creep at higher stresses. Microstructural observations along with parametric variations of creep rates were useful in identifying the underlying deformation mechanisms.}, number={10}, journal={ACTA MATERIALIA}, author={Gollapudi, S. and Charit, I. and Murty, K. L.}, year={2008}, month={Jun}, pages={2406–2419} }
 @article{gollapudi_bhosle_charit_murty_2008, title={Newtonian viscous creep in Ti-3Al-2.5V}, volume={88}, ISSN={["1478-6435"]}, DOI={10.1080/14786430802144154}, abstractNote={Biaxial creep tests were performed on fine-grained Ti–3Al–2.5V tubing at 823 and 873 K in the stress range σ/E  = 1.7  × 10−4 to σ/E  = 5.9  × 10−4. Subsequently, the creep data were analysed to determine the stress exponent and activation energy. A stress exponent value of 1 and an activation energy equal to that for grain boundary diffusion were suggestive of a Coble creep-controlled deformation regime. However, discrepancy between the experimental creep rates and Coble creep model predictions along with subsequent observation of deformed microstructures decorated with slip bands implied the operation of a different viscous creep mechanism. A slip band model proposed by Spingarn and Nix was found to provide a better description of the experimental strain rates rather than the conventional viscous creep mechanisms. High-resolution transmission electron microscopy studies confirmed the nature of these bands.}, number={9}, journal={PHILOSOPHICAL MAGAZINE}, author={Gollapudi, Srikant and Bhosle, Vikram and Charit, Indrajit and Murty, K. Linga}, year={2008}, pages={1357–1367} }
 @article{murty_charit_2008, title={Static strain aging and dislocation-impurity interactions in irradiated mild steel}, volume={382}, ISSN={["0022-3115"]}, DOI={10.1016/j.jnucmat.2008.08.008}, abstractNote={Interactions between dislocations and interstitial impurity atoms lead to strain aging phenomenon in ferritic steels that are affected by the defects produced during neutron radiation exposure. We present here results on static strain aging in a silicon-killed mild steel before and after neutron irradiation. It is noted that the degree of strain aging (as measured by the yield point following restraining) decreased with increasing neutron dose resulting in essentially non-aging type at the highest dose (∼1019 n/cm2). The strain aging kinetics were investigated using data at various aging temperatures and were found to be unaffected by the neutron radiation exposure. These experimental results are compared to those observed in dry hydrogen treated (partially denitrided) samples and are correlated with models on Cottrell locking.}, number={2-3}, journal={JOURNAL OF NUCLEAR MATERIALS}, author={Murty, K. L. and Charit, I.}, year={2008}, month={Dec}, pages={217–222} }
 @article{murty_charit_2008, title={Structural materials for Gen-IV nuclear reactors: Challenges and opportunities}, volume={383}, ISSN={["0022-3115"]}, DOI={10.1016/j.jnucmat.2008.08.044}, abstractNote={Generation-IV reactor design concepts envisioned thus far cater toward a common goal of providing safer, longer lasting, proliferation-resistant and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-IV reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This paper presents a summary of various Gen-IV reactor concepts, with emphasis on the structural materials issues depending on the specific application areas. This paper also discusses the challenges involved in using the existing materials under both service and off-normal conditions. Tasks become increasingly complex due to the operation of various fundamental phenomena like radiation-induced segregation, radiation-enhanced diffusion, precipitation, interactions between impurity elements and radiation-produced defects, swelling, helium generation and so forth. Further, high temperature capability (e.g. creep properties) of these materials is a critical, performance-limiting factor. It is demonstrated that novel alloy and microstructural design approaches coupled with new materials processing and fabrication techniques may mitigate the challenges, and the optimum system performance may be achieved under much demanding conditions.}, number={1-2}, journal={JOURNAL OF NUCLEAR MATERIALS}, author={Murty, K. L. and Charit, I.}, year={2008}, month={Dec}, pages={189–195} }
 @article{srikant_marple_charit_murty_2007, title={Characterization of stress rupture behavior of cornmercial-purity-Ti via burst testing}, volume={463}, ISSN={["0921-5093"]}, DOI={10.1016/j.msea.2006.06.145}, abstractNote={Abstract An understanding of the stress rupture behavior of Ti alloy tubing is of primary importance for structural applications in energy technology. The stress rupture properties were evaluated using burst testing of closed-end, thin-walled tubing at varied test temperatures and internal pressures. The rupture data are correlated using the Larson–Miller parameter. The uniform hoop strains were also measured along with rupture times from which the strain-rates were calculated. These results were fitted to Monkman–Grant relation with the aim of extrapolating to in-service stress levels. The activation energy for creep deformation was calculated from the Arrhenius equation, and the experimental data were analyzed using Dorn parameters. The analysis indicated a transition from a power-law controlled creep regime to power-law breakdown at high stresses. Transmission electron microscopy studies corroborated the transition in mechanism from power-law region to a power-law breakdown region.}, number={1-2}, journal={MATERIALS SCIENCE AND ENGINEERING A-STRUCTURAL MATERIALS PROPERTIES MICROSTRUCTURE AND PROCESSING}, author={Srikant, G. and Marple, B. and Charit, I. and Murty, K. L.}, year={2007}, month={Aug}, pages={203–207} }
 @article{charit_seok_murty_2007, title={Synergistic effects of interstitial impurities and radiation defects on mechanical characteristics of ferritic steels}, volume={361}, ISSN={["0022-3115"]}, DOI={10.1016/j.jnucmat.2006.12.003}, abstractNote={Ferritic steels are generally used in pressure vessels and various reactor support structures in light water reactors. They are known to exhibit radiation embrittlement in terms of decreased toughness and increased ductile–brittle transition temperature as a result of exposure to neutron radiation. The superimposed effects of strain aging due to interstitial impurity atoms on radiation embrittlement were considered first by Wechsler, Hall and others. Here we summarize some of our efforts on the investigation of synergistic effects between interstitial impurity atoms (IIAs) and radiation-induced point defects, which result in interesting effects at appropriate temperature and strain rate conditions. Two materials, a mild steel and a pressure vessel steel (A516 Gr.70), are evaluated using tensile and three-point bend tests.}, number={2-3}, journal={JOURNAL OF NUCLEAR MATERIALS}, author={Charit, I. and Seok, C. S. and Murty, K. L.}, year={2007}, month={Apr}, pages={262–273} }
 @misc{murty_charit_2006, title={Texture development and anisotropic deformation of zircaloys}, volume={48}, ISSN={["0149-1970"]}, DOI={10.1016/j.pnucene.2005.09.011}, abstractNote={This paper is a review of the texture development in zirconium alloys (in the form of thick walled tube reduced extrusion or TREX, thin-walled tubing and sheets) of importance to light and heavy water nuclear reactor technology along with the resultant anisotropic mechanical properties. Quantitative characterization of texture and mechanical anisotropy are emphasized leading to procedures useful to fabricators in optimizing textures for good formability as well as for acceptable in-service performance. A brief history of the development of zirconium alloys is presented followed by texture development and characterization. Mechanical anisotropy is discussed in terms of transverse contractile strain ratios from which the formability (B parameter) is derived. Results on the effect of annealing temperature as well as test temperature on anisotropy parameters are presented. The review concludes with a brief summary of texture effects on creep, stress corrosion cracking and hydride formation. Recent advances in fuel cladding bring out the challenges in characterizing the texture and anisotropy due to Nb additions and microstructural gradients in the new Zircaloys™, 1 such as Zirlo™, 2, Duplex™, 3 and Triclad ™, 4.}, number={4}, journal={PROGRESS IN NUCLEAR ENERGY}, author={Murty, KL and Charit, I}, year={2006}, pages={325–359} }
 @book{murty_charit, title={An introduction to nuclear materials: Fundamentals and applications}, publisher={Weinheim, Germany: Wiley-VCH}, author={Murty, K. L. and Charit, I.} }
 @inproceedings{gollapudi_bhosle_charit_murty, title={Low stress viscous creep in a ti-3al-2.5v tubing under internal pressurization}, booktitle={TMS 2010 139th Annual Meeting & Exhibition - Supplemental Proceedings, vol 1: Materials processing and properties}, author={Gollapudi, S. and Bhosle, V. and Charit, I. and Murty, K. L.}, pages={755–762} }