[en] Data on human reliability in control rooms indicate that human reliability is unacceptably low. This is particularly important under difficult unexpected situations in emergencies where the operator's deteriorated performance may lead to irreversible hazardous processes in the plant. Today, based on extensive research on the role of human element in technological systems, it is known that human error can not totally be eliminated in a modern flexible, changing environment, such as nuclear power plant, by conventional style designs. In this paper the innovative concept of error tolerance has been further explored to be utilized in supporting the cognitive functions of operators during the emergencies in nuclear power plants

[en] Application of probabilistic safety assessment to evaluate the safety of hazardous facilities will be fulfilled when the results have been processed meaningfully. The purpose of the importance analysis is to identify major contributors to core damage frequency that may include accident initiators, system failures, component failures and human errors. In this paper, Fussell-Vesely measure of importance was applied to the results of probabilistic safety assessment study of Tehran Research Reactor. This analysis is done using systems analysis programs for hands-on integrated reliability evaluations software

[en] Probabilistic safety assessment application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. In this document the application of the probabilistic safety assessment to the Tehran Research Reactor is presented. The level 1 practicabilities safety assessment application involved: Familiarization with the plant, selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantifications, fault tree constructions and quantification, human reliability, component failure data base development and dependent failure analysis. Each of the steps of the analysis given above is discussed with highlights from the selected results. Quantification of the constructed models is done using systems analysis programs for hands-on integrated reliability evaluations software

[en] In this paper, a Simulation Environment (SE) is introduced which has the capability of performing comprehensive and accurate analysis of different types of nuclear reactors and behaviour of their fuel under various steady-state and accident conditions. The SE currently consists of three main novel cods, a CFD-based, multidimensional code which works in transient and steady states for both single and two-phase flows and two separate neutronic codes. First code solves the point kinetics equations with multi-group delayed neutron precursors using four different methods. The second code solves multidimensional multi-group diffusion equations in transient and steady states and isotropic or anisotropic forms. Using the finite volume method for spatial discretization has provided considerable flexibility regarding geometry and boundary conditions, so that the most sophisticated details of fuels geometry such as spacer grids could be modelled accurately. The most important innovations incorporated into this SE are ''simultaneous solution'' approach which is a novel method for coupling the neutronic and thermohydraulic analyses based on the use of the point-implicit solver, ^multi-region porous media'' model which is an improvement over the original model by modifying the method of resistance coefficients evaluation and the approach by which this model is applied to the core of nuclear reactor and an introduction to the ''improved SIDK'', a temporal discretization method for accident analysis which has been obtained by combing the key feature of the original method that is decoupling of the prompt and delayed neutron equations with the separated solution approach of the point-implicit solver. The advantages provided by the mentioned innovations are partially indicated by presenting some practical examples and to demonstrate the accident modelling capabilities of this novel SE some relevant results associated with the case studies on TRR are presented and discussed which includes analyses of two DBAs i.e. RIA and LOCA. (author)

[en] Short communication

[en] The trip of one reactor coolant pump accident is simulated with using the IAEA-developed WWER-1000 simulation computer software (IAEA-Code) in this article. The results are compared with the information presented for Bushehr Nuclear Power Plant for the same scenario (analyzed by the code Dynamika-97). The obtained results are compatible with the Bushehr Nuclear Power Plant's given data and show a good overall agreement between them

[en] According to the incident/accident reports, unsuccessful implementation of steam generator feed-and-bleed procedure is one of the most important events in nuclear power plants operation which greatly contributes to the level of risk of the plants. Generally, the loss of all feed water pumps flow (as one of the precursors) results in failure to maintain adequate cooling of the reactor core unless the operating crew initiate and follow the feed-and-bleed procedure correctly and timely. In this paper, firstly, a Human Reliability Analysis (HRA) event tree is presented delineating the major human activities and errors in the implementation of the steam generator (SG) feed-and-bleed procedure following the loss of (both normal and emergency) water feed to four SGs of Bushehr Nuclear Power Plant unit1 (BNPP-1). Secondly, the graphical method of task analysis as a part of HRA is used as a means of delineating correct and incorrect human actions. To be used in the probabilistic risk assessment (PRA), the outputs of the HRA event trees are fed into the system event trees, functional event trees or system fault trees. As a part of a probabilistic risk assessment of BNPP-1 and to assess the reliability of control room operators, a human reliability analysis model is applied based on the THERP (Technique for Human Error Rate Prediction) technique. The THERP method is used in the form of event trees named as the probability tree diagrams. In this research the Human Reliability Analysis event tree is constructed based on the background information and assumptions made and on a similar NPP task analysis. It is done so because the BNPP-1 is not an operational nuclear power plant. Thirdly, based on NUREG/CR-1278 Handbook, a computer program has been developed in Visual Basic language and used to illustrate the major human activities and determination of error rates of operators in the course of the implementation of the steam generator feed-and-bleed procedure. Finally, total failure rate of BNPP-1 control room operators in the relevant steps of ''immediate Actions'' and ''Follow-up Actions'' is determined in the framework of a team work in which a modified concept of dependence among the control room operators is used. (author)

[en] A structural integrity assessment program has been developed at the National Nuclear Safety Department in cooperation with Amir Kabir University of Technology. The program is mainly based on the ASME XI [1] and R6 [2] codes. The program has four major modules: flaw assessment evaluation based on R6 Option 1, flaw assessment based on ASME XI, fatigue crack growth evaluation, and stress corrosion cracking evaluation. In this paper, the modules of the program are described briefly, and a benchmarking example is presented

[en] The deficient safety culture (S.C) is the center of safety issues of nuclear industry. To benefit from the advantages of nuclear technology and considering the fact of potential hazards of accidents in nuclear installations it is essential to view safety as the highest priority. S.C is an amalgamation of values, standards, morals and norms of acceptable behavior. Organizations having effective S.C show constant commitment to safety as a top level priority. Furthermore, the personnel of a nuclear facility shall recognize the safety significance of their tasks. Many people even those who work in the field of safety do not have a correct understanding of what S.C looks like in practical sense. In this study, by conducting a survey according to IAEA-TECDOC-1329 in some nuclear facilities, the S.C within the Iranian nuclear facilities is assessed. The human and organizational factors in Tehran Research Reactor are evaluated using a questionnaire method with active participation of the reactor operators. The results sho w that the operators are pretty aware of the subject. Also it has been identified some areas of improvement. (authors)

[en] The analysis of the past major NPPs accidents, TMI, Chernobyl and Fukushima Daiichi shows that causes of these accidents can be explained by a complex combination of human, technological and organizational factors. One of the findings of accident investigations and risk assessments is the growing recognition of the impact of cultural context of work practices on safety. The assumed link between culture and safety, epitomized through the concept of safety culture, has been the subject of extensive research in recent years. The term “safety culture” was first introduced into the nuclear industry by the IAEA in INSAG-1 to underline the role and importance of the organizational factors. The objective of this paper is to conduct an assessment of some safety culture indicators of Bushehr Nuclear Power Plant (BNPP-1).