[en] The traditional approach to engineering asset management, which primarily includes management and operation and maintenance activities, in a nuclear plant is based on engineering judgment. Here, the defence-in-depth and conservative criteria form the basis of decisions. Ideally, it is expected that the decisions should be based on performance of systems, structures and components (SSC), which should also take plant safety margin into consideration. Probabilistic methods are widely used as a compliment to deterministic methods in assessing safety of nuclear reactors. A risk-based operations and maintenance management system has been developed for a research reactor. The results of a level-1 PSA study in terms of minimal cut-set equations along with other engineering plant information forms major input for this system. The system is expected to be operational in plant control room conditions in support of routine decisions. This paper presents one of the main applications of the plant Level-1 Probabilistic Safety Assessment, developed from plant specific data i.e. a risk monitor. The risk monitor has several applications that support in decision making. It is expected that the use of this system in real-time scenario will provide improved understanding of safety on one hand while enabling enhancement of plant economics on the other. Some the case studies are discussed here. (author)

[en] During 1994, a small water leak was observed from the upper aluminium thermal shield of Cirus reactor. Detailed investigations revealed that the leakage was from the weld joint of one of the 1 1/4 inch NB Sch. 80 coolant inlet pipes connected to the upper aluminium thermal shield. The location of the leak was identified by monitoring the stabilised water level in the vertical inlet pipe under stagnant condition. The exact location was identified by installing an inflatable seal arrangement inside the leaky pipe and inflating the seal at different elevations to isolate the leaky location and ensuring that the leak was completely stopped. This location was about 15 feet below the operating floor of the reactor. The pipe was visually inspected with the help of a fibre-scope to assess the condition of the inner surface. Eddy current testing was also carried out for volumetric examination. This revealed one more localised flaw on the outer surface little above the leaky joint. A hollow plug, with expandable rings, having C-shaped cross section at both the ends and a straight portion in the middle to cover the defective region, was developed and qualified in a mock-up station after extensive trials. In view of the site constraints, a flexible hollow link assembly was engineered, for installing the plug remotely. The inner surface of the pipe was cleaned using an emery brush and a deburring tool. The plug was then installed covering the leak area and the rings were expanded by remote tightening. The shield was hydro-tested satisfactorily. (author)

[en] Sea water is used as the ultimate heat sink for Cirus and Dhruva reactors. Sea water is pumped through a common header to both the reactors using vertical turbine pumps located in a 40 feet deep chamber. A traveling water screen with 10 mm mesh is provided in the chamber at the upstream side of the pump. This screen is rotated at specific intervals with washing water jet supply on, for cleaning the screen of any entrapped matter. During the monsoon in year 2004, on one occasion, while the screen was being operated, its shear pin got broken. On inspection it was found that few baskets had got bent, preventing its movement along the haulage chain links through the guide brackets. The bent baskets were in almost completely choked condition with large collection of plastic sheets and it appeared that they got bent due to the resultant high liquid differential pressure across it. The matter was investigated in detail. This paper deals with the observations, conclusions and remedial measures w.r.t the bending incident of the traveling water screen baskets. (author)

[en] Efficient and reliable operation of secondary coolant pumps in a nuclear reactor is very important as heat generated in the reactor core is transferred through heat exchangers to the ultimate heat sink, with the help of these pumps. At Dhruva, (A 100 MWt natural uranium fuelled, nuclear research reactor), the reactor core is cooled by heavy water, the primary coolant in a closed loop, which, in turn, is cooled by secondary coolant (Process water). The secondary coolant is recirculated in closed loop by a set of single stage, double suction centrifugal pumps operating in parallel. Frequent bearing failures at pump driven end (DE side), pump-motor coupling failures, pump- seal leakages etc. were observed. These failures were considered, initially, to be of random nature. A close condition monitoring programme, undertaken subsequently, could identify the existence of severe cavitations in one of the pumps due to various reasons like increased wear ring clearances, changes in operating condition of the pumps etc. This paper describes the observations and various methods employed for identifying the root cause for frequent failures of secondary coolant pumps. By addressing the root cause, the conditions promoting failures are either eliminated or reduced to minimum. (author)

[en] One of the key elements that determines the operational safety of a Nuclear Reactor is the training and technical knowledge of its operators about the behaviour of their plant under normal, transient and accidental conditions. The operation of a complex system like a nuclear reactor requires thoroughly trained and licensed manpower. Simulators greatly support in the training and transfer of knowledge to the plant operators by making use of the latest available technologies. Simulator based training facilitate the trainees to learn the particular module and repeat till he gets mastery. It enhances the training quality and performance of the trainee. Full scope, non-replica Simulator for 100 MW Dhruva research reactor is indigenously designed, commissioned and made operational in BARC in co-ordination with ECIL. Apart from normal training, important Postulated Initiating Events (PIEs) like Loss of Offsite Power, Loss of Coolant Accident, Loss of Regulation Incident, etc. have been modelled. This information will provide an insight to verify and validate the plant Emergency Operating Procedures (EOPs). The simulator will also be used for development of human factor models as well as the design of future VDU based control rooms. This will also enable the operators to align themselves from the traditional panel based human-machine interface (HMI) to personal computers. Hard wired graphic panels were reproduced in the form of soft touch screens. The development activities were completed in a time bound manner. The development contract was awarded to a vendor (ECIL) with whom the requisite expertise and indigenously developed simulation software was available. The relevant data on design and operation parameters, details of interlocks, alarm and trip settings, Process and Instrumentation (P and I) diagrams, etc. pertaining to each system was provided by reactor group for model development. (author)

[en] Reactivity meter finds application in Control Rod worth evaluation, status of core reactivity measurements, calibration experiments of reactivity devices and worth estimation of fuel rods and irradiated samples. Reactivity can be calculated using Inverse Kinetics equations at any power level. The source strength can be assumed as zero for reactivity calculation in a critical reactor core and a core without any external neutron source. However, the source strength needs to be determined for reactivity calculation at lower power levels and for subcritical reactors. This report discusses reactivity measurement using real time computation based on Inverse Kinetics Method at sufficiently high power level of a critical reactor and reactivity changes during rod drop in the reactor. The studied algorithm was tested on simulated data as well as real time data from reactor and consistent results were found. For subcritical region of reactor operation, a code has been developed based on the literature available to estimate reactivity and source strength using power profile of the reactor. (author)

[en] The industrial workplace is a dynamic environment where various activities, including machinery operations and material handling, take place simultaneously. While this nature of the industry has numerous economic benefits, and same time it presents potential fire hazards that can cause catastrophic destruction if not appropriately addressed. A fire hazard may include a situation that increases the likelihood of a fire or may impede escape in the event a fire occurs. Fire safety is the set of practices intended to reduce destruction caused by fire. Fire safety measures include those that are intended to prevent the ignition of an uncontrolled fire and those that are used to limit the spread and impact of a fire. In the realm of safety protocols,. Accidents, whether in residential, commercial, or industrial settings, pose significant threats to life, property, and the environment. Fire emergencies, in particular, demand swift and efficient responses to mitigate potential devastation. Fire drills stand as indispensable tools in preparing individuals to respond effectively to fire-related incidents and mitigate the fire at its origin thus preventing catastrophic destruction. Simulated exercises are designed to ensure that individuals and organizations are well-prepared to respond promptly and effectively in the event of a fire emergency. This comprehensive analysis delves into the various ways in which fire drills contribute to accident prevention, emphasizing their significance in fostering a culture of safety and readiness

[en] Quality Assurance (QA) in safety culture represents a strategic framework within organizations that aims to ensure the highest standards of safety and are consistently met and improved upon. In high-risk radioactive industries, the integration of QA practices into safety culture is imperative for preventing accidents, protecting the workforce, and maintaining compliance with stringent regulations. The integration of QA principles ensures that safety measures are not only reactive but also proactive; fostering a culture of continuous improvement, risk mitigation and vision for zero incident. This paper explores the multifaceted role of Quality Assurance in safety culture, examining its significance, challenges, and impact on organizational performance

[en] Reliability based methods have been widely used for the safety assessment of plant system, structures and components. These methods provide a quantitative estimation of system reliability but do not give insight into the failure mechanism. Understanding the failure mechanism is a must to avoid the recurrence of the events and enhancement of the system reliability. Root cause analysis provides a tool for gaining detailed insights into the causes of failure of component with particular attention to the identification of fault in component design, operation, surveillance, maintenance, training, procedures and policies which must be improved to prevent repetition of incidents. Root cause analysis also helps in developing Probabilistic Safety Analysis models. A probabilistic precursor study provides a complement to the root cause analysis approach in event analysis by focusing on how an event might have developed adversely. This paper discusses the root cause analysis methodologies and their application in the specific case studies for enhancement of system reliability. (author)

[en] Reactivity meter is a device to measure reactivity instant to instant using real time computation based on inverse kinetics equation (IK). It finds application in control rod worth evaluation, calibration of reactivity devices and indication of point of criticality during reactor start up. To calculate reactivity in real-time specifically for Apsara-U (in the regime where source term can be neglected), a code has been developed based on IK equations using the power profile from neutron detectors. It has been tested for the real time data from Apsara-U. Calculated reactivity values of CSR-1 and 2 drop from 6 detectors using the power profile are shown here. In Apsara-U, photoneutron source occur in the BeO reflector region and the same can be accounted for measuring reactivity in the sub-critical region (reactor going sub-critical from critical state) as proposed by Tamura. Further, for measuring reactivity when reactor goes to subcritical state from another subcritical state, in case of unknown source, the new subcriticality and source term (photoneutron source contribution in this case) can be estimated as given in. The two mentioned type of calculation have been done and validated for simulated data. It will be extended for real time data