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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.

Cellular polymer foams find extensive applications as energy absorbers under static, impact and blast loads due to their capacity to absorb energy under constant stress up to full densification strain (Ashby et al. Metal foams: a design guide. Butterworth-Heinemann, Oxford, 2000 [1]; Gibson and Ashby, Cellular solids: structure and properties. Cambridge University Press, Cambridge, 1999 [2]). In the present paper, the material characterization of natural latex cellular polymer foam used in the field air blast experiments (Venkataramana et al. Proc Eng 173:547–554, 2017 [3]; Venkataramana et al. Numerical simulation of blast wave mitigation using foam impregnated with water. DAE BRNS Symposium on Multiscale Modeling of Materials and Devices (MMMD-2014), 2014 [4]) performed to assess the blast mitigation effect of fluid-filled open-cell pin-core polymer foam is presented. Static compression tests were performed to study the quasi-static behavior of the pin-core latex polymer foam. The Ogden hyper-elastic material parameters are determined from the analysis of static compression test data using ADINA (automatic dynamic incremental nonlinear analysis) v 9.2, ADINA R&D, Inc., Watertown [5]). Further, results of the static compression tests on dry foam and water-saturated foam are compared and discussed. In addition, the dynamic behavior of the pin-core latex cellular foam is investigated by conducting drop tower impact experiment and numerical simulation of the drop tower experiment using LS-DYNA (LS-DYNA v971, Livermore Software Technology Corporation, Livermore [6]). The stress–time history from the drop tower experiment is compared with that of simulation, and good correlation is found between the two sets of results.

Fracture studies were carried out on 170 mm OD welded straight pipes having circumferential through-wall crack in the weld under quasi-cyclic loading. The specimens were made of Type 304 LN stainless steel and welded using conventional Shielded Metal Arc (SMA) welding or Narrow Gap (NG) welding. The specimens were subjected to different amplitudes of load/displacement, which was decided based on the monotonic load carrying capacity of the specimens. It was found that the failure load reduces to 78.5% and 74.0% of the monotonic failure load for resisting 50 and 100 loading cycles, respectively in the case of NG welded pipes. The NG welded specimens showed more energy absorbing capability and crack growth resistance, when compared with SMA welded specimens. Crack growth was more stable in the case of specimens tested under displacement control; whereas the same was unstable in the case of load-controlled experiments and specimens failed by sudden collapse.

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 third stage of Indian nuclear power programme envisages the use of thorium as the fertile material with 233U, which would be obtained from the operation of Pu/Th based fast reactors in the later part of the second stage. Thorium-based reactors have been designed in many configurations, from light water-cooled designs to high-temperature liquid metal-cooled options. Another option, which holds promise, is the molten salt-fuelled reactor, which can be configured to give significant breeding ratios. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian molten salt breeder reactor (IMSBR). Presently, various design options and possibilities are being studied from the point of view of reactor physics and thermal hydraulic design. In parallel, fundamental studies on natural circulation and corrosion behaviour of various molten salts have also been initiated.

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.

India is currently operating two BWR built by General Electric Company. The design features of these reactors are similar to the Fukushima's BWR except some better containment features in Indian BWR. This paper discusses the enveloping scenario of station blackout of infinite duration with no operator action and no component failure. The paper describes the details of modelling the TAPS-BWR plant model including SCDAP modelling of reactor core in system code RELAP5 and further thermal hydraulic safety assessment of station blackout scenario. The analysis brought out effectively the response of the plant to this high-pressure severe accident scenario. The time line of the severe accident progression will give details of various stages of accident progression along with hydrogen generation, which will be useful in evolving suitable severe accident management guidelines.

In-service inspections of many nuclear power plants have revealed that cracks are most likely to occur in a weld or in the regions near the weld. In part-I, the problem of a stationary crack lying at the centre of a weld in a pure bending specimen SE(PB) was analysed. The detailed structure of the global plastic fields for a deep crack, under fully plastic condition, was presented. Aspects related to the state of stress at the base-weld interface were discussed. To enhance our understanding of the weld strength mismatch effects, other commonly used fracture specimens, that is, compact tension C(T) and middle tension M(T) specimens having a weld centre crack were analysed in the present investigation. The influence of weld mismatch on the structure of global stress fields (leading to plastic yielding of the ligament) as well as on the crack tip constraint was studied. It is demonstrated that, when a crack is postulated at the centre of a weld, a family of stress fields proposed in part-I for a SE(PB) specimen is applicable to a C(T) specimen also. The studies performed in this article, along with part-I, have established that in comparison to slip line field analysis, the modified upper bound theorem is simple and more general. It is applicable to macroscopically homogeneous materials and can also account for weld mismatch effects.

Currently leak-before-break based design and fracture stability assessment considers the earthquake as once applied non-cyclic load, which monotonically increases up to its maximum magnitude. The cyclic character of earthquake load and associated cyclic tearing failure mode (tearing–fatigue regime) are not explicitly considered while demonstrating stability of a through wall cracked pipe. To single out and quantify the impact of cyclic tearing on fracture stability of circumferential through wall cracked pipes, a systematic programme involving extensive number of tests is carried out. The load reduction factors (β) or safety margins, required to ensure cracked pipe stability for specified number of reversible load cycles, are directly evaluated from cyclic and corresponding monotonic pipe tests. They are presented as Cyclic Tearing Failure Assessment Diagram (CTFAD) and a β-factor equation. The proposed β-factors when applied on the monotonic critical load of a cracked pipe (evaluated using existing procedures/methods) will ensure its fracture integrity for specified number of load cycles. The safety margin required against cyclic fracture instability under different levels of earthquakes is presented.

Recent trend in risk informed and risk based approaches in life management issues have certainly put the focus on developing estimation methods for real risk. Idea of employing risk as an optimising measure for in-service inspection, termed as risk based inspection, was accepted in principle from late 80s. While applying risk based inspection, consequence of failure from each component needs to be assessed. Consequence evaluation in a Process Plant is a crucial task. It may be noted that, in general, the number of components to be considered for life management is very large and hence the consequence evaluation resulting from their failures (individually) is a laborious task. Screening of critical components is usually carried out using simplified qualitative approach, which primarily uses influence factors for categorisation. This necessitates logical formulation of influence factors and their ranges with a suitable technical basis for acceptance from regulators. This paper describes application of risk based inspection for H2S based Process Plant along with the approach devised for handling the influence factor related to the quantity of H2S released.

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.

The objective of the present study is to understand the Fatigue Crack Growth (FCG) behavior in austenitic stainless steel pipe welds. The tests have been carried out on Compact Tension (CT) specimens machined from the actual narrow gap pipe welds. The notch was located in Weld Centre Line (WCL), Heat Affected Zone (HAZ), HAZ-Fusion Line (FL) interface and the parent metal. The comparisons have been made for the FCG rate for different locations of the crack. The tests were carried out using standard procedure of ASTM E647. The results of FCG rate of narrow gap welds (HAZ, WCL, HAZ-FL interface and parent material) have been compared with that of conventional V-groove welds of Shielded Metal Arc Welding (SMAW) process. The comparison shows that the fatigue crack growth resistance of narrow gap welds is better than conventional groove welds. The FCG resistance of weld and parent material is better than HAZ location in the Paris regime. However, parent material shows superior threshold crack driving force range.

In geophysical survey system, a Time Domain Electro-Magnetic (TDEM) technique is used to detect the presence of uranium in the earth crust. For heliborne survey, certain equipments, such as power converter and data acquisition system are installed on a helicopter. The power converter is responsible for the generation of high current pulses to feed the transmitter for setting up the required electromagnetic field. The power converter consists of a free running inverter, which sits over a DC-DC converter. Data acquisition system acquires data from various sensors for further processing and interpretation. Data acquisition system consists of a highly vibration sensitive onboard computer (IDAN boxes) rigidly mounted over a plate inside the data acquisition system enclosure. Since the power converter and data acquisition system are installed on the floor inside the surveying helicopter, it is mandatory to qualify these equipments for vibration & shock as per MIL standard before onboard installation. To evaluate the structural integrity of power converter and data acquisition system under various loading conditions of vibration and shock, a preliminary estimation of stresses and deflections was made by hand calculations and then, finite element analysis has been performed to optimize the design. Modal analysis has been performed to estimate the natural frequencies of power converter. Since the natural frequencies of the power converter were within the disturbing frequency range (i.e., 10 to 84 Hz), thickness of converter base plate and inverter base plate has been increased to avoid resonance. Since the data acquisition system consists of a highly vibration sensitive onboard computer, it has been proposed to mount it on isolators to minimize the vibration. To evaluate the structural integrity of data acquisition system under various loading conditions, finite element analysis has been performed. This paper presents the methodology adopted for design and analysis of the equipments for the specified input of vibration and shock.

The proposed heavy water moderated and light water cooled pressure tube type boiling water reactor works on natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the main heat transport system configuration. These multiple (four) interconnected loops influence the steam drum level control adversely through the common reactor inlet header. Alternate design studies made earlier for efficient control of SD levels have shown favorable results. This has lead to explore further the present scheme with the compartmentalization of CRIH into four compartments catering to four loops separately. The conventional 3-element level control has been found to be working satisfactorily. The interconnections between ECCS header and inlet header compartments have also increased the safety margin for various LOCA and design basis events. The paper deals with the SD level control aspects for this novel MHT configuration which has been analyzed for various PIEs (Postulated Initiating Events) and found to be satisfactory.

Pipe bends or elbows are commonly used components for nuclear power plant piping system. In service, these piping components are subjected to internal pressure in addition to bending loads and the internal pressure is known to have a significant effect on the load carrying capacity of these components. Hence, a systematic study was carried out to investigate and quantify the effect of internal pressure on the fracture behaviour of elbows used in nuclear power plant piping system. Fracture studies were conducted on five 219 mm diameter carbon steel elbows with and without internal pressure under in-plane opening moment. The investigations have shown that the presence of a circumferential notch at the intrados has a more detrimental effect on the fracture behaviour of the elbow, when compared with the presence of an axial notch at the intrados. It is also found that internal pressure plays a significant role in reducing the ovalization.

The overall goal of nuclear power plant safety is to protect individuals, society and the environment from undue radiological hazard so that nuclear power production does not significantly add to the health risks to which individuals and society are already exposed. This paper addresses the safety principles followed during the design phase of life cycle of a nuclear power plant. The principles followed such as safety classification, design rules based on failure modes, detailed stress analysis, stress categorization, consideration of design basis events, failure probability, flaw tolerance, leak-before-break are described. Engineering structures always contain flaws, albeit of very small size. Fatigue and fracture are the two important failure modes affected by flaws. Thus flaw tolerance becomes very important. This is assessed by applying fracture mechanics principles. The R6 procedure, which is used for evaluation of structures containing flaws, has been incorporated in the software BARC-R6. Improvements by way of shell-nozzle junction pull-out, adoption of hot wire GTAW with narrow gap technique have been brought out. Post Fukushima incidence, resistance to seismic loading and containment design have assumed great importance. The paper describes these aspects in detail. Regulatory aspects of seismic design regarding siting, Seismic margin assessment, base isolation, retrofitting are the aspects covered under seismic design. Under the action of seismic loading, the piping in a nuclear power plant piping is vulnerable to a phenomenon called ratcheting. The process of seismic margin assessment and consideration of ratchetting has been backed up by a large experimental data. The experiments carried out on structures and piping components form a part of the paper.

Following the Fukushima accident, the safety features of Nuclear Power Plants (NPP) are being re-examined worldwide including India to demonstrate capabilities to cope with severe accidents. In order to restore public confidence and support for nuclear power, it is felt necessary to design future NPPs with near zero impact outside the plant boundary and thus enabling elimination of emergency planning in public domain. Authors have identified a set of safety features which are needed to be incorporated in advanced reactors to achieve this goal. These features enabling prevention, termination, mitigation and containment of radioactivity for beyond design basis accidents arising from extreme natural events are essential for achieving the goal of elimination of emergency planning in public domain. Inherent safety characteristics, passive and engineered safety features to achieve these functions are discussed in this paper. Present trends and future developments in this direction are also described briefly.

Removal of decay heat from an operating reactor during a prolonged station blackout condition is a big concern for reactor designers, especially after the recent Fukushima accident. In the case of a prolonged station blackout condition, heat removal is possible only by passive means since no pumps or active systems are available. Keeping this in mind, the AHWR has been designed with many passive safety features. One of them is a passive means of removing decay heat with the help of Isolation Condensers (ICs) which are submerged in a big water pool called the Gravity Driven Water Pool (GDWP). The ICs have many tubes in which the steam, generated by the reactor core due to the decay heat, flows and condenses by rejecting the heat into the water pool. After condensation, the condensate falls back into the steam drum of the reactor. The GDWP tank holds a large amount of water, about 8000 m3, which is located at a higher elevation than the steam drum of the reactor in order to promote natural circulation. Due to the recent Fukushima type accidents, it has been a concern to understand and evaluate the capability of the ICs to remove decay heat for a prolonged period without escalating fuel sheath temperature. In view of this, an analysis has been performed for decay heat removal characteristics over several days of an AHWR by ICs. The computer code RELAP5/MOD3.2 was used for this purpose. Results indicate that the ICs can remove the decay heat for more than 10 days without causing any bulk boiling in the GDWP. After that, decay heat can be removed for more than 40 days by boiling off the pool inventory. The pressure inside the containment does not exceed the design pressure even after 10 days by condensation of steam generated from the GDWP on the walls of containment and on the Passive Containment Cooling System (PCCS) tubes. If venting is carried out after this period, the decay heat can be removed for more than 50 days without exceeding the design limits.

The main objective of this work is to investigate the sloshing behavior in a baffled and unbaffled three dimensional annular-sectored water pool (i.e., tank) which is located at dome region of the primary containment. Initially two case studies were performed for validation. In these case studies, the theoretical and experimental results were compared with numerical results and good agreement was found. After the validation of present numerical procedure, an annular-sectored water pool has been taken for numerical investigation. One sector is taken for analysis from the eight sectored water pool. The free surface is captured by Volume of Fluid (VOF) technique and the fluid portion is solved by finite volume method while the structure portions are solved by finite element approach. Baffled and un-baffled cases were compared to show the reduction in wave height under excitation. The complex mechanical interaction between the fluid and pool wall deformation is simulated using a partitioned strong fluid-structure coupling.

The paper aims at demonstration of reduction in residual stress, distortion and sensitisation in austenitic stainless steel pipe welds. The welding processes considered are Gas Tungsten Arc Welding (GTAW) and Shielded Metal Arc welding (SMAW) along with Hot wire GTAW with narrow gap welding technique. It was shown that residual stress is significantly reduced in case of hot wire GTAW. The reduction in residual stress is due to the low heat input and high deposition rate. Lower heat input leads to higher cooling rate giving more margins on sensitization in heat affected zone (HAZ). This has been confirmed by measuring temperature and cooling rate during welding in HAZ of the weld joints. Susceptibility to sensitization of the welds has been compared by carrying out ASTM E262 Practice A and E along with Electo-Potentiokinetic Reactivation (EPR) method. The results of the tests show that values are within acceptable limit for both the cases. However, hot wire GTAW has marginally better sensitization resistance. Fatigue crack growth rate and fracture resistance behavior of the above mentioned weld joints have been compared by carrying out tests on the specimens (Compact Tension) and full scale pipe welds with crack. Fatigue crack growth rate and fracture toughness of the weld joints (hot wire GATW) is superior to conventional GTAW and SMAW. The paper presents the details of the above mentioned studies.

Advanced and innovative reactor systems are considering the use of large pools as heat sink for various safety functions like decay heat removal or containment cooling. These designs generally have heat exchangers immersed in the pool. For enhanced safety and reliability, preferred heat transfer mode is considered to be passive so that heat sink availability is maintained even in failure of power supply and active components. However, heat transfer by natural convection in large pools poses a problem of thermal stratification. As a result of natural convection, hot layers of water may accumulate over the relatively cold one and in turn inhibit the natural convection itself. Not only the heat transfer performance may get deteriorated but some structural parts of the pool like concrete wall may be subjected to high temperature which is not desirable. In this paper, a new concept of employing shrouds around the heat source is proposed. These shrouds provide multiple natural circulation loops around the heat source, thereby facilitating mixing of hot and cold fluid, which eliminate stratification. The concept has been applied to the Gravity Driven Water Pool (GDWP) of Advance Heavy Water Reactor (AHWR) in which Isolation Condensers (ICs) tubes are submerged for decay heat removal of AHWR using ICS and thermal stratification phenomenon was predicted without and with ICS. Results indicate that the shrouds have application in elimination of thermal stratification in GDWP.

The integrity assessment of the primary piping components needs to be demonstrated under normal operation cyclic loadings as well as under complex cyclic loadings of extreme magnitude as may come during a severe earthquake event. In order to understand material's cyclic plasticity and fatigue-ratcheting behaviour, systematic experimental and analytical investigations have been carried out on specimens of SA333Gr.6 carbon steel and SS304LN stainless steel. The material specification of SA333Gr.6 is same as used in Primary Heat Transport (PHT) piping of Pressurized Heavy Water Reactors (PHWRs) and material specification of SS304LN steel is same as proposed for Indian Advance Heavy Water Reactor's (AHWR's) Main Heat Transport (MHT) piping. The test program included the tensile, axial fatigue and uniaxial ratcheting tests to establish the material's mechanical properties and cyclic plasticity behavior. The results of these tests have been investigated in details using few popular finite element cyclic plasticity models to understand and quantify the material's cyclic plasticity behavior. The studies revealed the need to modify the Chaboche model to simulate the LCF/cyclic plasticity and ratcheting under different strain/stress amplitudes loading conditions. On accounting for modification, the Chaboche model nicely predicted the LCF and ratcheting response for all the tests. The tests, finite element analyses result and their interpretations have been presented in this paper.

The paper presents experimental and numerical work carried out on a full-scale four storey reinforced concrete (RC) structure for seismic assessment by pushover method. For practicality, a portion of an existing structure having certain eccentricities was replicated for the experimental setup. The structure was detailed as per non-seismic reinforcement detailing norms of Indian Standards. The experiment was carried out as a round robin exercise, in which various institutes in India participated and presented pre-test results in the form of pushover curves. A large variation in the pre-test results highlighted that the result of a pushover analysis is highly sensitive to the adopted modeling techniques. This paper reports the details and results of the experiment and focuses on the need of modeling various structural nonlinearities to obtain realistic results. The results of pre-test analysis by various research groups, in which the emphasis was given on modeling issues as well as a more efficient post-test numerical procedure is also presented and compared. It is shown that a basic pushover analysis considering only flexural failures may not be able to achieve a realistic simulation, thus it is mandatory to develop relatively simple, yet effective models to consider more complex phenomena such as joint shear failures, to achieve realistic predictions.

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.

A reinforced concrete (RC) framed structure detailed according to non-seismic detailing provisions as per Indian Standard was tested on shake table under dynamic loads. The structure had 3 main storeys and an additional storey to simulate the footing to plinth level. In plan the structure was symmetric with 2 bays in each direction. In order to optimize the information obtained from the tests, tests were planned in three different stages. In the first stage, tests were done with masonry infill panels in one direction to obtain information on the stiffness increase due to addition of infill panels. In second stage, the infills were removed and tests were conducted on the structure without and with tuned liquid dampers (TLD) on the roof of the structure to investigate the effect of TLD on seismic response of the structure. In the third stage, tests were conducted on bare frame structure under biaxial time histories with gradually increasing peak ground acceleration (PGA) till failure. The simulated earthquakes represented low, moderate and severe seismic ground motions. The effects of masonry infill panels on dynamic characteristics of the structure, effectiveness of TLD in reducing the seismic response of structure and the failure patterns of non-seismically detailed structures, are clearly brought out. Details of design and similitude are also discussed.

The design of Nuclear Power Plant piping is based on Leak-Before-Break (LBB) design philosophy. The integrity of cracked nuclear piping system in LBB is often examined from conventional fracture mechanics. Conventionally the fracture analysis is performed by assuming infinite compliance of the piping system, which means that the stiffening effect of the connected uncracked piping system is fully neglected. However, in reality the ends of the pipe are connected to the rigid pressure vessels, primary coolant pipes, etc. In order to study the effect of compliance on the fracture behaviour of pipes, fracture studies were carried out on fourteen specimens of 219 mm outer diameter carbon steel pipes having circumferential through-wall crack. The fracture tests were conducted under four point bending. Out of the fourteen specimens, twelve specimens were tested in fixed end condition (finite compliance) and the remaining two specimens were tested in simply supported condition (infinite compliance). These pipes were made of SA 333 Gr. 6 carbon steel. Extensive instrumentation was carried out on these pipes to collect various data, e.g., load, load-line displacement (LLD), crack mouth opening displacement (CMOD), rotation at the ends (flanges), deflection at critical locations and surface crack growth. Surface crack growth was monitored using Image Processing Technique (IPT). The investigations have shown that the load carrying capacity of the pipes under finite compliance condition (fixed end condition) increases when compared to infinite compliance (simply supported condition). The load carrying capacity of the pipes under finite compliance weakly depends on the initial crack angle where as the load carrying capacity of the pipes under infinite compliance strongly depends on the initial crack angle. The safety margin is quite large particularly with a stiff piping system. This benefit in terms of increase in safety margin from LBB point of view is due to the moment redistribution. This paper discusses the effects of notch angle, span and the support conditions on the fracture behaviour of pipes with circumferential through-wall crack.

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.

A systematic experimental and analytical study has been carried out to investigate the effect of vibration loading on the fatigue life of the piping components. Three Point bend (TPB) specimens machined from the actual pipe have been used for the evaluation of Paris constants by carrying out the experiments under vibration + cyclic and cyclic loading as per the ASTM Standard E647. These constants have been used for the prediction of the fatigue life of the pipe having part-through notch of a/t = 0.25 and aspect ratio (2c/a) of 10. Predicted results have shown the reduction in fatigue life of the notched pipe subjected to vibration + cyclic loading by 50% compared to that of cyclic loading. Predicted results have been validated by carrying out the full-scale pipe (with part-through notch) tests. Notched pipes were subjected to loading conditions such that the initial stress-intensity factor remains same as that of TPB specimen. Experimental results of the full-scale pipe tests under vibration + cyclic loading has shown the reduction in fatigue life by 70% compared to that of cyclic loading. Fractographic examination of the fracture surface of the tested specimens subjected to vibration + cyclic loading have shown higher presence of brittle phases such as martensite (in the form of isolated planar facets) and secondary micro cracks. This could be the reason for the reduction of fatigue life in pipe subjected to vibration + cyclic loading.

The objective of the present study is to understand the fatigue crack growth behavior in austenitic stainless steel pipes and pipe welds by carrying out analysis/predictions and experiments. The Paris law has been used for the prediction of fatigue crack growth life. To carry out the analysis, Paris constants have been determined for pipe (base) and pipe weld materials by using Compact Tension (CT) specimens machined from the actual pipe/pipe weld. Analyses have been carried out to predict the fatigue crack growth life of the austenitic stainless steel pipes/pipes welds having part through cracks on the outer surface. In the analyses, Stress Intensity Factors ( K ) have been evaluated through two different schemes. The first scheme considers the ‘ K ’ evaluations at two points of the crack front i.e. maximum crack depth and crack tip at the outer surface. The second scheme accounts for the area averaged root mean square stress intensity factor ( K RMS ) at deepest and surface points. Crack growth and the crack shape with loading cycles have been evaluated. In order to validate the analytical procedure/results, experiments have been carried out on full scale pipe and pipe welds with part through circumferential crack. Fatigue crack growth life evaluated using both schemes have been compared with experimental results. Use of stress intensity factor ( K RMS ) evaluated using second scheme gives better fatigue crack growth life prediction compared to that of first scheme. Fatigue crack growth in pipe weld (Gas Tungsten Arc Welding) can be predicted well using Paris constants of base material but prediction is non-conservative for pipe weld (Shielded Metal Arc Welding). Further, predictions using fatigue crack growth rate curve of ASME produces conservative results for pipe and GTAW pipe welds and comparable results for SMAW pipe welds.

A new nuclear power reactor under design study is a vertical pressure tube type boiling light water cooled and heavy water moderated. One of the passive design features of this reactor is the heat removal through natural circulation of primary coolant at all power level with no primary coolant pumps. Nuclear plants are mainly base load units, but the proposed plant with various advance features has to operate in load following mode i.e. Reactor follows Turbine (in a limited range). In this mode, any alteration in turbine load results in the steam pressure change. The steam pressure error is fed to the Reactor Regulating System (RRS), which changes the reactor power to control the system pressure. To study this mode of plant operation, a plant simulation model with the feedbacks from various controllers has been developed using the RELAP5 code. This integrated plant model has been used for simulating the load-varying scenario for a change in plant load. All the process dynamics, modeling, design verification and performance issues are discussed in this paper.

Pushover analysis has gained significant popularity as an analytical tool for realistic determination of the inelastic behaviour of RC structures. Though significant work has been done to evaluate the demands realistically, the evaluation of capacity and realistic failure modes has taken a back seat. In order to throw light on the inelastic behaviour and capacity evaluation for the RC framed structures, a 3D Reinforced concrete frame structure was tested under monotonically increasing lateral pushover loads, in a parabolic pattern, till failure. The structure consisted of three storeys and had 2 bays along the two orthogonal directions. The structure was gradually pushed in small increments of load and the corresponding displacements were monitored continuously, leading to a pushover curve for the structure as a result of the test along with other relevant information such as strains on reinforcement bars at critical locations, failure modes etc. The major failure modes were observed as flexural failure of beams and columns, torsional failure of transverse beams and joint shear failure. The analysis of the structure was by considering all these failure modes. In order to have a comparison, the analysis was performed as three different cases. In one case, only the flexural hinges were modelled for critical locations in beams and columns; in second the torsional hinges for transverse beams were included in the analysis and in the third case, joint shear hinges were also included in the analysis. It is shown that modelling and capturing all the failure modes is practically possible and such an analysis can provide the realistic insight into the behaviour of the structure.

In level-3 Leak-Before-Break (LBB) analysis, stability of a postulated through-wall circumferential crack is demonstrated by simplified fracture mechanics calculations. Detailed experimental studies, conducted by the authors, have revealed that the conventional assessment procedure used to demonstrate LBB is too conservative. There is a large factor of safety due to system indeterminacy. It was observed that the critical load of a cracked piping system (with even a large through-wall circumferential crack of about 120°) is of the order of 75–90% of the collapse load of the uncracked piping system. Reduction in load carrying capacity is even less for a piping system having an off-centre crack. This article discusses the above-mentioned aspects in detail. Detailed 3-D elastic-plastic finite element analyses of some of these tests were performed. The suitability of these numerical results to predict crack initiation load in light of the experimental data is discussed.

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.

A combined mechanical property evaluation methodology with ABI (Automated Ball Indentation) simulation and Artificial Neural Network (ANN) analysis is evolved to evaluate the mechanical properties for Carbon Manganese Steel (SA-333 Grade-6) and Stainless Steel (SS-304LN). The experimental load deflection data is converted into meaningful mechanical properties for these materials and their evaluated property is verified with experimental tensile specimen results. An ANN database is generated with the help of contact type finite element analysis by numerically simulating the ABI process for various magnitudes of yield strength (σ yp ) (200 MPa–400 MPa) with a range of strain hardening exponent (n) (0.05–0.5) and strength coefficient (K) (600 MPa–1600 MPa). For the present problem, a ball indenter of 1.57 mm diameter having Young’s modulus higher than test piece is used to minimize the error due to indenter deformation. Test piece dimension is kept large enough in comparison to the indenter configuration in the simulation to minimize the deflection at the outer edge of the test piece. Further, this database after the neural network training; is used to analyse measured material properties of different test pieces. The ANN predictions are reconfirmed with contact type finite element analysis for an arbitrary selected test sample. The methodology evolved in this work can be extended to predict material properties for any irradiated nuclear material in the service. Extensions of the ABI tests and the associated database analysis could lead to evaluation of the indentation energy to fracture needed for the structural integrity assessment of aged components.

In water-cooled nuclear power reactors, significant quantities of hydrogen could be produced following a postulated loss-of-coolant accident (LOCA) along with nonavailability of emergency core cooling system (ECCS). Passive autocatalytic recombiners (PAR) are implemented in the containment of water-cooled power reactors to mitigate the risk of hydrogen combustion. In the presence of hydrogen with available oxygen, a catalytic reaction occurs spontaneously at the catalyst surfaces below conventional ignition concentration limits and temperature and even in presence of steam. Heat of reaction produces natural convection flow through the enclosure and promotes mixing in the containment. For the assessment of the PAR performance in terms of maximum temperature of catalyst surface and outlet hydrogen concentration an in-house 3D CFD model has been developed. The code has been used to study the mechanism of catalytic recombination and has been tested for two literature-quoted experiments.

In this work, Automated Ball Indentation (ABI) technique is based on load controlled multiple indentations (at a single penetration location) of a polished surface by a spherical indenter (0.7 to 1.46 mm) and indentation depth is progressively increased to a maximum user specified limit with intermediate partial unloading. This technique permits measurement of yield strength, stress-strain curve, strength coefficient and strain hardening exponent. ABI Testing was carried out on samples of Zr-Nb2.5 (Pressure Tube Material) with different heat treatment conditions in which temperature was varying (550 degree to 900 degree and retention time was varying 0.5 to 6 hour and furnace cooled. For all these test material and conditions, the ABI derived results were in very good agreement with those from conventional standard test methods.

SUMMARY Smart materials such as shape memory alloys (SMAs) are recently being used in earthquake engineering applications to control the response of structures. In this paper, a shape-memory alloy damper device made up of austenite wires (e.g. Nickel Titanium wires) is used as a passive energy absorber. NiTi wires are very attractive for passive vibration control as they have a pseudo-elastic property and can sustain large amounts (upto 10% strain) of inelastic deformation. Moreover, in contrast to regular metallic materials they can recover that deformation. The damper device is designed, fabricated and tested. Validity study is made using a thermo-mechanical model of SMA taking into account the residual martensite accumulation irreversibly due to cyclic forward/reverse martensitic transformation. Further an analytical study of a supporting steel structure of dump tanks with and without SMA dampers subjected to design wave earthquake loading is carried out. Performance of the structure with SMA dampers is compared with that of the same structure with yielding dampers. Copyright © 2010 John Wiley & Sons, Ltd.

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 presents the methodology, which will be adopted, for quantifying the effect of uncertainties on the peak clad temperature of an advanced natural circulation nuclear reactor. The method relies on probabilistic analysis, treating uncertain parameters as random variables. The paper will cover a case of a loss-of-coolant accident due to a 200% (that is, double ended) break of the largest pipe with partial unavailability of the low-pressure emergency core cooling system. The break has been postulated at the inlet header, which is the largest pipe in the main heat transport system. For this assessment a two-step procedure has been adopted. In the first step the probability of the peak clad temperature exceeding 800°C has been evaluated using the response surface, generated from the results of thermal-hydraulic analyses. One of the fuel failure criteria for this reactor is the peak clad temperature exceeding 800°C. Such a high temperature is expected during typical large-break loss-of-coolant accident conditions. The thermal-hydraulic analyses, using the computer code RELAP5/MOD3.2, were done for several cases involving different combinations of six selected uncertain parameters. The probabilistic analysis was carried out using Monte Carlo and first-order reliability methods. The first step results in conditional probability of the peak clad temperature exceeding the criteria subject to the condition of a 200% break in the inlet header. The probability of a 200% break is calculated in the second step. The probability of an inlet header pipe weld rupture has been evaluated based on probabilistic fracture assessment. The pipe break analysis considers the uncertainties in strength, fracture, and stress corrosion properties and initial crack/flaw sizes produced during fabrication or welding. It also accounts pre-service and in-service inspection, inspection quality, and different damage mechanisms such as fatigue and intergranular stress corrosion cracking. The combined results of both these steps give the overall probability of the peak clad temperature exceeding 800°C.

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.

The use of thorium is necessary from a long-term objective of sustainability of nuclear energy resources in India. Its use requires reprocessing to separate the fissile 233U for sustained operation using 232Th–233U. Over the years, owing to sustained technology developmental initiatives, India has gained significant experience in all aspects of the thorium fuel cycle. In addition to briefly explaining advantages and challenges of thorium fuel cycle, this paper covers the Indian experiences in this direction. It also briefly describes the designs of Indian reactors based on thorium fuel cycle, for example, the Advanced Heavy Water Reactor (AHWR), as well as conceptual designs of proposed future reactor systems.

Rational seismic design procedures necessitate comprehensive evaluation of nuclear piping systems under large amplitude seismic loads. This comprehensive assessment requires accurate prediction of inelastic response of piping system till failure to ensure adequate margins for unexpected beyond design basis events. The present paper describes the details of experimental and numerical studies of inelastic response of pressurized piping system under seismic loading. Shake table test has been carried out on a three-dimensional stainless steel piping system under internal pressure and seismic load. The amplitude of base excitation has been increased till failure of the piping system. The tested piping system has been analyzed using iterative response spectrum (IRS) method for various levels of excitation. The comparison of numerical and experimental results is given in the paper.

This paper describes the experimental studies carried out on cracked austenitic stainless steel pipe and pipe welds under bending loads. Pipe welds were produced by gas tungsten arc welding (GTAW) and shielded metal arc welding (SMAW). Fracture resistance curves for pipe and pipe welds were compared. Results indicate that the fracture resistance of pipe and pipe weld (GTAW) is comparable but that of pipe weld (GTAW+SMAW) is inferior. Cracks do not deviate from their original plane during propagation as observed in the cases of carbon steel pipe and pipe welds. The fracture resistance of pipe welds does not depend on the loading histories to which it has been subjected prior to fracture test. Initiation and crack propagation were observed prior to the maximum moment. An existing limit load expression is applicable for the pipe base material but gives non-conservative results for the pipe welds. Multiplication factors have been suggested for the pipe welds for evaluation of limit loads using the existing expression. Fracture resistance for the pipe and compact tension specimens have also been compared for base material and welds.

In this paper, factors for calculating the J-resistance curve from the measured crack mouth opening displacement (CMOD) record are derived for circumferential through wall cracked straight pipes subjected to pure bending. Rigorous elastic plastic finite element analyses of around 125 circumferential through wall cracked straight pipes subjected to four point bending have been carried out to validate the proposed equations. The factors have also been validated with a few fracture experimental results on straight pipes. For these tests the J–R curve has been evaluated using the conventional as well as using the proposed factors along with the moment-CMOD record.

The current Leak Before Break (LBB) assessment is based primarily on the monotonic fracture tearing instability. In it the maximum design accident load is compared with the fracture-tearing resistance load. The effect of cyclic loading has not been generally considered in the fracture assessment of nuclear power plant piping. It is a well-known fact that reversible cyclic loading decreases the fracture resistance of the material, which leads to increased crack growth. Indian nuclear power reactors consider Operational-Basis-Earthquake (OBE) and Safe-Shutdown-Earthquake (SSE) events in the design of various structures, systems, and components. Keeping this in view a series of cyclic tearing tests have been conducted on straight pipes, made of ASTM SA333 Gr.6 carbon steel. This is the material of primary heat transport (PHT) piping material of Indian Pressurized Heavy Water Reactors (PHWR). In this series 13 tests have been carried out circumferentially through wall cracked seamless and circumferential seam welded straight pipes under reversible cyclic bending loading. All the tests have been conducted under quasistatic, i.e., slow loading rates and dynamic inertia effects are not considered. The cyclic test results have been compared with the corresponding monotonic pipe fracture test results. These test results and its comparison with corresponding monotonic tearing clearly illustrate the need of addressing the reduction in apparent fracture toughness of material under reversible cyclic loading and the safe number of load cycles in the LBB assessment.

Earthquake load, which is cyclic in nature and of short duration, is the main design basis accident load for designing the primary heat transport (PHT) piping components of Nuclear Power Plants. Adequate protection of piping components from the effects of earthquake requires detailed knowledge of strength and deformation characteristics of the components and assemblies making up the piping system. Fracture behaviour of 23 pipes was studied by conducting cyclic tests under four-point bending. Fifteen pipes were of carbon steel (SA3 3 3 Grade 6) of 219 mm, 324 mm and 406 mm ODs and eight pipes were of stainless steel (AISI Type 304LN) of 168 mm OD with through-wall circumferential crack. Six pipes were tested under displacement control and the rest under load control. The effect of various parameters such as location and size of initial crack in base metal and weld metal, cyclic load range and displacement increment on crack growth and number of cycles for failure was investigated. The investigations showed significant reduction in the fracture resistance under cyclic loading conditions. Crack growth equations have been proposed for carbon steel pipes with and without weld under cyclic tearing.

Weld joints of SA 333 Gr. 6 steel pipe have been prepared using GTAW root pass and SMAW filling of V-groove according to welding procedure specifications (WPS) conforming to ASME Sec IX, a procedure commonly used to fabricate nuclear piping system components. The objective of the present study was to characterise the base metal, weld deposit and HAZ of the pipe weld in terms of their chemical composition and metallurgical, mechanical and fracture mechanics properties. The fracture toughness behaviour of the welds and HAZ has been characterised by J-integral parameters. The fatigue crack growth rate has been characterised by Paris Law. The stretched zone width (SZW) at the fracture surface has been measured with SEM to evaluate the initiation fracture toughness. The values of initiation fracture toughness estimated on the basis of SZW and blunting line from EGF recommendation have been compared. The fracture mechanics properties of the base metal, weld and HAZ have been compared and correlated. Fra...