[en] 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 η_pl 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

[en] The integrity of fuel rods under normal, abnormal and accident conditions is an important consideration during fuel design of advanced nuclear reactors. The fuel matrix and the sheath form the first barrier to prevent the release of radioactive materials into the primary coolant. An understanding of the fuel and clad behaviour under different reactor conditions, particularly under the beyond-design-basis accident scenario leading to large scale transients, is always desirable to assess the inherent safety margins in fuel pin design and to plan for the mitigation the consequences of accidents, if any. The severe accident conditions are typically characterized by the energy deposition rates far exceeding the heat removal capability of the reactor coolant system. This may lead to the clad failure due to fission gas pressure at high temperature, large- scale pellet-clad interaction and clad melting. The fuel rod performance is affected by many interdependent complex phenomena involving extremely complex material behaviour. The versatile experimental database available in this area has led to the development of powerful analytical tools to characterize fuel under extreme scenarios

[en] The performance of high burnup fuel is affected by a number of phenomena, such as, conductivity degradation, modified radial flux profile, fission gas release from high burnup structures, PCMI, burnup dependent thermo-mechanical properties, etc. The modelling strategies of some of these phenomena are available in literature. These can be readily incorporated in a fuel modelling performance code. The computer code FAIR has been developed in BARC over the years to evaluate the fuel performance at extended burnup and modelling of the fuel rods for advanced fuel cycles. The present paper deals with the high burnup issues in the fuel pins, their modelling strategies and results of the case studies specifically involving high burnup fuel. (author)

[en] Integrity assessment of piping components with postulated cracks is important for safe and reliable operation of power plants. While various equations and methods are available for prediction of the load bearing capacity of pipes and elbows, it is very important to choose the correct equation and method whose predictions are consistent, safe but not too conservative with respect to the experimental results. Towards this goal, a comprehensive Component Integrity Assessment Program was initiated under a joint MPA-BARC collaborative program where a large number of austenitic and ferritic pipes and elbows of nominal diameter of 50-400 mm with various crack configurations and sizes were tested. These test results along with results of previous tests were analysed with various available limit load equations present and also with the R6 method. Based on the comparison of these test results and predictions, the correct equation and method are recommended to reliably predict the load bearing capacity of flawed pipes and elbows. Integrity assessment, Piping, Leak-before-break (LBB) behaviour, Analytical/numerical and experimental investigations, Engineering and fracture mechanics methods, Limit load calculations

[en] In the fuel bundles of Indian PHWRs, the differential thermal expansion of pins leads to loading of the endplate. The varying axial gap between the pellets in different pins affects the pin expansion. Thus the startup and shutdown of the reactor causes variation in the stresses. Detailed analysis of endplate is carried out to estimate allowable fatigue cycles assuming statistical distribution of axial gaps. (author)

[en] The use of conventional elements to model singular point leads to slow convergence. The use of singularity elements in place of conventional elements has helped to improve convergence and accuracy. There are a number of singular elements available in the literature. However, the most popular one is the quarter point 8-noded degenerate quadrilateral element. This element satisfies all the three convergence requirements - the rigid body, the interelement continuity and the constant strain conditions. The most severe limitation of this element is its ability to model only a square root singularity. The other important element is the 3-noded variable order singular element. This has the ability to model variable order singularities. In one of our recent publications, we have derived the shape functions of a 6-noded variable order singular element. However, both of these variable order singular elements do not satisfy constant strain criteria and cannot be used under thermal loads. The present formulation may be viewed as a further modification to that formulation in this direction

[en] Pipe coupling using clamps are becoming more and more popular components in nuclear power plants and are used in place of conventional flanges because of their compactness, easy maintenance and more reliability. They are used in large numbers in Pressurised Heavy Water Reactos (PHWR) such as at the joint between feeder pipe and End-fitting, in F/M housing etc. Integrity of these clamps have direct effect on overall safety of the nuclear power plants. This necessitates proper design, fabrication, installation and maintenance of these components. Proper design of these clamps is a challenge to the designer. This is because of changing boundary conditions at the interface with the hub during various stages of loading. A detail stress analysis of clamps considering changing boundary conditions under various loading can be done using finite element technique. In the following sections, two finite element modelling methods to simulate clamps along with hubs are described. Both these methods assumed absence of friction between the clamp and hubs during bolting, whereas absence of relative movement between them was assumed during other stages of loadings

[en] The concept of partial thickness method developed by Marcal and King (1967) has been used for elasto-plastic analysis of plane stress/plane strain and axisymmetric bodies by finite element method. A computer algorithm developed on the above basis has been checked by solving three typical problems, namely, elastic-plastic deformation of: (1) a thick cylindrical shell, (2) perforated tension strip under inplane loa.ding, and (3) notched tension strip. The results have been compared with other known solutions/experimental results. (M.G.B.)

[en] One of the most appealing features of the Pressurised Heavy Water Reactors is the online refuelling capability. For this a fuelling machine is used. This machine opens a reactor channel by removing a seal plug and a shield plug and then does the necessary fuelling by pushing fuel bundles from a fuel magazine by rams. After necessary fuelling the machine closes the channel automatically. One of the most important parts of the fuelling machine is its pressure housing which becomes a part of the reactor channel during refuelling operation. It houses the fuel magazine, separators and rams. Beside channel pressure and other mechanical loads, the pressure housing experiences thermal transients during refuelling. The housing consists of two cylindrical shells having one end-closer in each. They are connected with each other by a large sized coupling. There are many holes on both the end-closers to accommodate ram movement, separators and magazine drive mechanisms. Some of these holes intersect with each other in the housing end-closures and hence end-closures are reinforced accordingly. This also makes the end-closures nonsymmetric. In the following sections the various analysis done to compute general stress distribution, stress concentration factors near to various holes, temperature transients during refuelling and also allowable fatigue cycles for pressure housing of fuelling machine for the proposed 500 MWe are described

[en] Leak-Before-Break (LBB) is being used to design the primary heat transport piping system of 500 MWe Indian Pressurized Heavy Water Reactors (IPHWR). The work is categorized in three directions to demonstrate three levels of safety against sudden catastrophic break. Level 1 is inherent in the design procedure of piping system as per ASME Sec.III with a well defined factor of safety. Level 2 consists of fatigue crack growth study of a postulated part-through flaw at the inside surface of pipes. Level 3 is stability analysis of a postulated leakage size flaw under the maximum credible loading condition. Developmental work related to demonstration of level 2 and level 3 confidence is described in this paper. In a case study on fatigue crack growth on PHT straight pipes for level 2, negligible crack growth is predicted for the life of the reactor. For level 3 analysis, the R6 method has been adopted. A database to evaluate SIF of elbows with throughwall flaws under combined internal pressure and bending moment has been generated to provide one of the inputs for R6 method. The methodology of safety assessment of elbow using R6 method has been demonstrated for a typical pump discharge elbow. In this analysis, limit load of the cracked elbow has been determined by carrying out elasto-plastic finite element analysis. The limit load results compared well with those given by Miller. However, it requires further study to give a general form of limit load solution. On the experimental front, a set of small diameter pipe fracture experiments have been carried out at room temperature and 300 degrees C. Two important observations of the experiments are - appreciable drop in maximum load at 300 degrees C in case of SS pipes and out-of-plane crack growth in case of CS pipes. Experimental load deflection curves are finally compared with five J-estimation schemes predictions. A material database of PHT piping materials is also being generated for use in LBB analysis