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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.

The recent Fukushima accident has resulted in an increased need for passive safety systems in upcoming advanced reactors. In order to enhance the global contribution and acceptability of nuclear energy, proven evidence is required to show that it is not only green but also safe, in case of extreme natural events. To achieve and establish this fact, we need to design, demonstrate and incorporate reliable ‘passive safety systems’ in our advanced reactor designs. In Nuclear Power Plants (NPPs), the use of passive safety systems such as accumulators, condensing and evaporative heat exchangers and gravity driven cooling systems provide enhanced safety and reliability. In addition, they eliminate the huge costs associated with the installation, maintenance and operation of active safety systems that require multiple pumps with independent and redundant electric power supplies. As a result, passive safety systems are preferred for numerous advanced reactor concepts. In current NPPs, passive safety systems which are not participating in day to day operation, are kept isolated, and require a signal and external energy source to open the valve. It is proposed to replace these valves by passive components and devices such as self-acting valves, rupture disks, etc. Some of these innovative passive valves, which do not require external power, have been recently designed, developed and tested at rated conditions. These valves are proposed to be used for various passive safety systems of an upcoming Nuclear Power Plant being designed by India. For example, the Hot Shutdown Passive Valves (HSPV), developed for the decay heat removal system keep the main heat transport system under hot conditions by passively sensing and controlling the system pressure. Another crucial and important valve which has been successfully developed is the Poison Injection Passive Valve (PIPV) for the Passive Poison Injection System. It not only provides higher reliability, but also ensures safe shutdown of the reactor in case of insider threats or malevolent acts in disabling active shutdown system of the reactor. Recently, an innovative valve called the Accumulator Isolation Passive Valve (AIPV) has been developed for the Emergency Core Cooling System (ECCS), which is engineered to mitigate the consequences of Loss of Coolant Accident (LOCA). During normal operation of the reactor, the pressurized accumulators (55 bar) are kept isolated from the reactor core (70 bar) by means of AIPVs. In case of a LOCA, these passive valves open when the main heat transport system pressure falls to a desired value. For prolonged cooling of the core, these passive valves regulate the discharge in a desired manner. These are non-standard, high pressure and high temperature valves, which are unavailable commercially and hence have to be indigenously designed and developed. This paper primarily deals with the design, development and testing of Accumulator Isolation Passive Valves (AIPV) proposed to be used in the ECCS. A 25 NB size AIPV has been designed and successfully tested at Integral Test Loop (ITL) under simulated reactor conditions. It is a self-acting, ANSI 600 rating valve, which requires no external energy (i.e., neither air nor electrical power). It not only provides passive isolation but also passively controls high pressure liquid discharge through it. The design concept of the valve, functional performance, in situ valve testing methodology and the test results at simulated conditions are discussed.

This paper addresses the strength and deformation capacity of stiff squat shear wall subjected to monotonic and pseudo-static cyclic loading using experiments and analysis. Reinforced concrete squat shear walls offer great potential for lateral load resistance and the failure mode of these shear walls is brittle shear mode. Shear strength of these shear walls depend strongly on softening of concrete struts in principal compression direction due to principal tension in other direction. In this work simulation of the behavior of a squat shear wall is accurately predicted by finite element modeling by incorporating the appropriate softening model in the program. Modification of model of concrete in the softening part is suggested and reduction factor given by Vecchio et al. (1994) is used in the model. The accuracy of modeling is confirmed by comparing the simulated response with experimental one. The crack pattern generated from the 3D model is compared with that obtained from experiments. The load deflection for monotonic loads is also obtained using softened truss theory and compared with experimental one.

To investigate several unresolved issues and to improve upon the existing equations for integrity assessment of pipes and pipe bends used in nuclear reactors, a comprehensive Component Integrity Test Program (CITP) was initiated at Bhabha Atomic Research Centre (BARC), India. As a part of this program, several fracture tests have been conducted on straight pipes and pipe bends, which forms a valuable data base. Simultaneously, analytical work has been undertaken to propose the improvements in the existing equations for optimized and more accurate integrity assessment of piping components. As an outcome of this analytical investigations, generalized equation of ηpl and γ have been proposed to evaluate J–R curve, study of transferability of fracture properties from specimen to component has been done to highlight the role of constraint parameter, new limit moment equations of elbows have been proposed and new J and COD estimation schemes of throughwall cracked elbows have been proposed. All these newly proposed equations have been experimentally validated with the test data generated under CITP.

Component failures due to excessive flow-induced vibration are still affecting the performance and reliability of nuclear power stations. Tube failures due to fretting-wear in nuclear steam generators, and vibration related damage of reactor internals are of particular concern. In the Indian nuclear industry, flow induced vibrations are assessed early in the design process and the results are incorporated in the design procedures. In this paper the details of flow induced vibration studies on internals like liquid zone control unit and poison injection units of heavy water filled calandria of 700 MWe Indian pressurized heavy water reactor is given. This includes computational fluid dynamics studies from which the velocities are extracted for the components lying inside the calandria. With these velocities as input, further studies are performed to predict the dynamic behavior of these components. Results of these calculations as well as conclusions derived from this investigation are presented. Based on the studies it has been established that flow induced vibration is not a concern in the present design of 700 MWe calandria internals.

The main function of the Steam Generator Pressure Control (SGPC) Program is to match the power (heat) generation in the reactor core with the heat removal in the steam generators (SGs). For most of the designs these programs have been over simplified to cater to the limitation of the instrumentation and control, hardware and software. The main objective of balancing the heat generation, transfer and removal gets lost in the process, which leads to reduction in the availability of the nuclear power plant. This is reflected in under utilization of the process and control system provisions to avoid reactor trips on low/high pressure. Most of the transients/accidents have their origin in the mismatch among the heat generated in the reactor core and the heat removal in the SGs. A new Advanced Process Control (APC) based supervisory controller is proposed to ride over the existing SGPC to achieve the goal. This APC makes use of the estimated/measured heat generation-removal error to alter the SGPC set point to tide over the transients after detection. The transients are detected based on the magnitude of this error to activate the APC. After tiding over the transient successfully the control switches back to the existing SGPC. For evaluation of this error additional instrumentation is proposed in the SG tube walls and other structures of the primary system. The Reactor Regulating System (RRS) which manipulates the reactor power, PHT pressure controller and the SGPC must work in a synchronized and optimized fashion so as to avoid a reactor trip following the PCP trip and other transients. With the existing design it is often observed that these control loops conflict with each other, as they are not designed with a multivariable integrated approach. The objective is being fulfilled by developing new supervisory APC logics to ride over the existing Proportional Integral Derivative (PID) controllers of various process loops. The new APC concept/logics formulation and supervisory control details along with performance tests are presented in this paper.

Every nuclear facility consists of various structures of varying importance. Standards and design guidelines recommend categorization of these structures based on their importance, safety function that they have to perform, hazard class, etc. In order to economize the designs, the structures that pose only industrial risk are recommended to be designed following the national practice. This paper presents the results of experiments performed on the beam-column joints of such structures to verify their seismic performances. The specimens were full-scale replicas of the joints of the original non safety related structures such as turbine buildings, intake structures and office buildings. The beam-column connections in the original structures either had non-seismic detailing or seismic detailing, but in order to have a comparison, every joint was tested with both types of detailing. The strength and ductility obtained from the experiments for both types was compared with the codal recommended values by different national standards. Conclusions and comments on the suggested values of allowable stresses, assumed ductility values and corresponding strength reduction factors have been made.

Passive energy dissipating devices like elasto-plastic dampers (EPDs) can be used for retrofitting of structures subjected to seismic loads. A model of reinforced concrete structure is tested on shake table with and without EPDs attached in its frames. Using a finite element model of the structure, linear and nonlinear time history analysis is carried out using Newmark's time integration technique. However, the most viable approach used by designers is response spectrum approach. Hence equivalent linearization techniques are used to address the nonlinearity of dampers and iterative response spectrum method is used for evaluating the response of the structure using equivalent damping and stiffness. The analytical maximum storey response of the structure is compared with experimental values and time history analysis values. It has been concluded that, iterative response spectrum technique using equivalent linearization techniques is simple and results in reasonably acceptable response of the structures retrofitted with energy dissipaters.

This paper summarizes the results from the investigations carried out on fiber reinforced concrete with steel fibers, lead fibers and a combination of the two (hybrid fibers). The intent of this research was to investigate the effect of the two types of fibers on mechanical and radiation shielding properties of concrete. Compressive strength, split tensile strength and flexural toughness were among the mechanical properties investigated and radiation shielding to gamma rays was investigated by comparing the attenuation provided by different types of concrete against each other and against blank readings without attenuation. The results clearly showed that the hybrid fibers showed a significant enhancement in both mechanical and radiation shielding properties.

This paper describes the results of fatigue studies on carbon steel piping materials and components of Indian Pressurized Heavy Water Reactors (PHWRs). The piping components include pipes and elbows, of outer diameter 219 mm, 324 mm and 406 mm, made of carbon steel (SA333 Gr.6 grade) material. Tests on actual pipes and elbows with part through notch were carried out to study the behaviour of crack growth under cyclic loading for different pipe sizes, notch aspect ratios, stress ratios, etc. During the tests, numbers of cycles for crack initiation from blunt notch were recorded with an accuracy of 0.1 mm. In conjunction with component tests, the experimental studies were also conducted on standard specimens to understand the effect of different variables such as size (thickness), type of specimen and components (elbow and pipe), welding, stress ratio, notch orientation on fatigue crack growth rate. The fatigue crack growth curve (da/dN versus ΔK) obtained from three-point bend specimen and pipe was compared with that given in ASME Section XI. The comparison shows that da/dN versus ΔK curves obtained from the specimen and pipe tests are nearly same. The analytical predictions for crack initiation and crack growth for the tested components were compared with experimental results. Such comparisons validate the modeling procedure for crack initiation and growth.

In Indian type pressurized heavy water reactors (PHWRs), two fuelling machines (F/Ms) are located on either side of the reactor. The behavior of the coolant channel and tie rod when fuelling machines are engaged on both the ends of the channel for refueling under earthquake is studied. It is observed that the loads acting on the coolant tube were high. In order to reduce the earthquake load acting on coolant channel the fuelling machines are restrained by energy absorbers called lead extrusion dampers (LED). The design of an LED is performed based on the extrusion principle. The aim of the paper is to study the effectiveness of such a device in reducing the response. It was observed that the stresses in the coolant channel due to earthquake are reduced significantly when LEDs are used. Response spectrum analysis is also carried out for the system with LED. In this analysis the LED is replaced with an equivalent viscous damper and stiffness obtained using Caughy’s linearisation techniques. Response spectrum analysis results and the nonlinear time history analysis results show good agreement.

The most commonly used method in the seismic analysis of structures is the response spectrum method. For seismic re-evaluation of existing facilities elastic response spectrum method cannot be used directly as large deformation above yield may be observed under Safe Shutdown Earthquake (SSE). The plastic deformation, i.e. hysteretic characteristics of various elements of the structure cause dissipation of energy. Hence the values of damping given by the code, which does not account hysteretic energy dissipation cannot be directly used. In this paper, appropriate damping values are evaluated for 5-storey, 10-storey and 15-storey shear beam structures, which deform beyond their yield limit. Linear elastic analysis is performed for the same structures using these damping values and the storey forces are compared with those obtained using inelastic time history analysis. A damping model, which relates ductility of the structure and damping, is developed. Using this damping model, a practical structure is analysed and results are compared with inelastic time history analysis and the comparison is found to be good.

Deterministic fracture analysis of cracked structures has been successful in demonstrating the behavior of structures under observed or postulated flaws. The key ingredients in a deterministic fracture mechanics analysis are the initial crack size, crack-driving force, applied stress and material properties. However, crack size, postulated accidental loading, mechanical and fracture parameters are often subjected to considerable scatter or uncertainty. Hence, the result of fracture mechanics analysis must be viewed with some skepticism. Quite often, conservative bounds on inputs are employed on these by providing a conservative estimate to fracture. However, it may lead to predict the overly conservative and unrealistic results. These uncertainties encourage us to adopt a statistical or probabilistic approach to the structural integrity. R6 is a well-known engineering assessment procedure for the evaluation of the structural integrity of the flawed structure. This paper presents the determination of failure probabilities of a straight pipe with a through-wall crack in the circumferential direction, using the R6 method. The pipe is subjected to an in-plane bending moment. The scatter in the crack size, mechanical and fracture properties are considered. The probability of failure is studied for variation in applied loads. Monte Carlo simulation is used to find the failure probability. The study is performed by constructing failure assessment diagram (FAD) using all the three options of R6. The results are compared with the approximate analytical formulations and with those available in the literature.