جستجوی مقالات مرتبط با کلیدواژه "psa" در نشریات گروه "فیزیک"

Success criteria analysis (SCA) is one type of deterministic analysis and in the case of the PSA level 1 framework, it is used in 3 areas. 1- determination of the end state for each sequence of the event tree, 2- deterministic confirmation of the fault tree analysis, 3- help to calculate the human error probability and identify its consequences. This research covers the 1st type of SCA. Success criteria as a deterministic analysis must be based on a realistic steady-state model of the plant, so the MELCOR code is used to model the plant which is a 2-loops Westinghouse PWR. When the steady-state model is achieved, the design basis accident i.e. MSLB is modeled. The event tree of the accident contains 12 sequences and it is needed to determine the end state for each sequence. Determining the end states through success criteria analysis is to express the exact condition of the NPP i.e., is it a safe end state or not, so two different criteria are used in order to identify the safe end state, the peak clad temperature must not exceed 1477 K or the pressure of the containment is not allowed to be increased more than the design pressure (0.4 Mpa). For mitigation of the MSLB accident consequences, two safety functions are required: 1- reactivity control and 2- heat removal by use of the main steam line isolation and emergency feed water in the secondary circuit. Both of the mentioned functions above are satisfied in the 1st sequence, but if there is a failure in the heat removal, the operator must start the primary feed and bleed process by use of the Safety Injection System (SIS) and Safety Depressurization Valve (SDV) in the primary circuit. According to the operation or failure of the required systems or functions, in some sequences, the above criteria are not met and it is not a safe end-state condition for the reactor. By implementing the success criteria analysis which is known as sequence analysis SCA, the end state for all sequences (except the 6th and 11th sequences that lead to Induced Steam Generator Tube Rupture (ISGTR) accident is determined.according to the results, none of the criteria above are violated in the 1st, 2nd, and 7th sequences and thus they lead to the safe end state. The design pressure criterion for the containment is not met in the 3rd and 8th sequences and there will be containment damage. The 4th, 5th, 9th, 10th, and 12th sequences lead to core damage as the peak clad temperature exceeds 1447 K

Safety, reliability and maintenance are key issues in the operation of nuclear power plants. Proper and organized implementation of maintenance activities can provide a platform for the safe operation of the power plant.In This research we try to increase the reliability of power plant systems and equipment by optimizing maintenance activities by introducing the reliability-based maintenance method (RCM).In This article the different approaches in the RCM program implemented in nuclear power plants and tries to choose an operational and efficient method in Bushehr nuclear power plant. Hence, the conflicts between classical RCM and simple RCM have been raised and an attempt has been made to resolve these conflicts through a comprehensive model.In this paper, two different RCM methods based on engineering judgment and statistical analysis have been used and the effect of these methods on heart rate has been evaluated. The method of combining FTA and recursive FMEA to evaluate equipment and select critical equipment according to the depth of analysis Optimally, used. In this method, by performing FTA and FMEA repeatedly at different levels of power plant analysis, critical equipment is identified using probabilistic safety assessment criteria