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[en] For licensing purposes, safety cases of Nuclear Power Plants (NPPs) must be presented at the Regulatory Authority with the necessary confidence on the models used to describe the plant safety behavior. In principle, this requires the repetition of a large number of model runs to account for the uncertainties inherent in the model description of the true plant behavior. The present paper propounds the use of bootstrapped Artificial Neural Networks (ANNs) for performing the numerous model output calculations needed for estimating safety margins with appropriate confidence intervals. Account is given both to the uncertainties inherent in the plant model and to those introduced by the ANN regression models used for performing the repeated safety parameter evaluations. The proposed framework of analysis is first illustrated with reference to a simple analytical model and then to the estimation of the safety margin on the maximum fuel cladding temperature reached during a complete group distribution header blockage scenario in a RBMK-1500 nuclear reactor. The results are compared with those obtained by a traditional parametric approach
[en] With the sustained development in computer technology, the use of more powerful computational tools becomes mandatory. The challenge today is to revisit safety features of the existing nuclear research reactors using new generation of computer tools. The objective is to verify that the safety requirements still met and when necessary to introduce some amendments coming from the new attainments. In the current paper the IAEA safety-related nuclear research reactors (RR) benchmark problem is reconsidered. The idea consists in performing static calculations of the benchmark using the last version of the MCNP5 code. This later offers updated code models and cross-section library. The results are afterwards compared with previous calculations and discussed
[en] Fuel management in PWR nuclear reactors is comprised of a collection of principles and practices required for the planning, scheduling, refueling, and safe operation of nuclear power plants to minimize the total plant and system energy costs to the extent possible. Despite remarkable advancements in optimization procedures, inherent complexities in nuclear reactor structure and strong inter-dependency among the fundamental parameters of the core make it necessary to evaluate the most efficient arrangement of the core. Several patterns have been presented so far to determine the best configuration of fuels in the reactor core by emphasis on minimizing the local power peaking factor (Pq). In this research, a new strategy for optimizing the fuel arrangements in a VVER-1000 reactor core is developed while lowering the Pq is considered as the main target. For this purpose, a Fuel Quality Factor, Z(r), served to depict the reactor core pattern. Mapping to ideal pattern is tracked over the optimization procedure in which the ideal pattern is prepared with considering the Z(r) constraints and their effects on flux and Pq uniformity. For finding the best configuration corresponding to the desired pattern, Cellular Automata (CA) is applied as a powerful and reliable tool on optimization procedure. To obtain the Z(r) constraints, the MCNP code was used and core calculations were performed by WIMS and CITATION codes. The results are compared with the predictions of a Neural Network as a smart optimization method, and the Final Safety Analysis Report (FSAR) as a reference proposed by the designer.
[en] A passive safety injection system (PSIS) is proposed for Chashma nuclear power plant-1 (CHASNUPP-1) type nuclear power plants, for the simplification of their safety systems. This system is based upon passive components and is proposed in place of the existing safety injection system, for safety enhancement. The functionality of the proposed system is analyzed using reactor simulation. For this purpose an intermediate size break LOCA is simulated using the simulation software APROS. For this transient, different thermal-hydraulic parameters of the proposed and other safety related systems are presented and discussed. The results obtained show that the proposed system works properly by performing its role in the transient, leading to cold shutdown conditions
[en] The trip setpoints for the reactor protection system of a 65-MWt advanced integral reactor have been analyzed through sensitivity evaluations by using the Transients and Setpoint Simulation/System-integrated Modular Reactor code. In the analysis, an inadvertent control rod withdrawal event has been considered as an initiating event because this event results in the worst consequences from the viewpoint of the minimum critical heat flux ratio and its consequences are considerably affected by the trip setpoints. Sensitivity evaluations have been performed by changing the trip setpoints for the ceiling of a variable overpower trip (VOPT) function and the pressure of a high pressurizer pressure trip function. Analysis results show that a VOPT function is an effective means to satisfy the acceptance criteria as the control rod rapidly withdraws: on the other hand, a high pressurizer pressure trip function is an essential measure to preserve the safety margin in the case of a slow withdrawal of the control rod because a reactor trip by a VOPT function does not occur in this case. It is also shown that the adoptions of 122.2% of the rated core power and 16.25 MPa as the trip setpoint for the ceiling of a VOPT function and the pressure of a high pressurizer pressure trip function are good selections to satisfy the acceptance criteria
[en] A decommissioning plan should be followed by a qualitative and quantitative safety assessment of it. The safety assessment of a decommissioning plan is applied to identify the potential (radiological and non-radiological) hazards and risks. Radiological and non-radiological hazards arise during decommissioning activities. The non-radiological or industrial hazards to which workers are subjected during a decommissioning and dismantling process may be greater than those experienced during an operational lifetime of a facility. Workers need to be protected by eliminating or reducing the radiological and non-radiological hazards that may arise during routine decommissioning activities and as well as during accidents. The risk assessment method was developed by using risk matrix and fuzzy inference logic, on the basis of the radiological and non-radiological hazards for a decommissioning safety of a nuclear facility. Fuzzy inference of radiological and non-radiological hazards performs a mapping from radiological and non-radiological hazards to risk matrix. Defuzzification of radiological and non-radiological hazards is the conversion of risk matrix and priorities to the maximum criterion method and the mean criterion method. In the end, a composite risk assessment methodology, to rank the risk level on radiological and non-radiological hazards of the decommissioning tasks and to prioritize on the risk level of the decommissioning tasks, by simultaneously combining radiological and non-radiological hazards, was developed.
[en] The simulation of nuclear power plant accident conditions requires three-dimensional (3D) modeling of the reactor core to ensure a realistic description of physical phenomena. The operational flexibility of Light Water Reactor (LWR) plants can be improved by utilizing accurate 3D coupled neutronics/thermal-hydraulics calculations for safety margins evaluations. There are certain requirements to the coupling of thermal-hydraulic system codes and neutron-kinetics codes that ought to be considered. The objective of these requirements is to provide accurate solutions in a reasonable amount of CPU time in coupled simulations of detailed operational transient and accident scenarios. These requirements are met by the development and implementation of six basic components of the coupling methodologies: ways of coupling (internal or external coupling); coupling approach (integration algorithm or parallel processing); spatial mesh overlays; coupled time-step algorithms; coupling numerics (explicit, semi-implicit and implicit schemes); and coupled convergence schemes. These principles of the coupled simulations are discussed in details along with the scientific issues associated with the development of appropriate neutron cross-section libraries for coupled code transient modeling. The current trends in LWR nuclear power generation and regulation as well as the design of next generation LWR reactor concepts along with the continuing computer technology progress stimulate further development of these coupled code systems. These efforts have been focused towards extending the analysis capabilities as well as refining the scale and level of detail of the coupling. This article analyses the coupled phenomena and modeling challenges on both global (assembly-wise) and local (pin-wise) levels. The issues related to the consistent qualification of coupled code systems as well as their application to different types of LWR transients are presented. Finally, the advances in numerical and computation techniques for coupled code simulations are summarized with outlining remaining challenges
[en] For the case where trains or channels of standby safety systems consisting of more than two redundant components are tested in a staggered manner, the standby safety components within a train can be tested simultaneously or consecutively. In this case, mixed testing schemes, staggered and non-staggered testing schemes, are used for testing the components. Approximate formulas, based on the basic parameter method, were developed for the estimation of the common cause failure (CCF) probabilities of the components under mixed testing schemes. The developed formulas were applied to the four redundant check valves of the auxiliary feed water system as a demonstration study for their appropriateness. For a comparison, we estimated the CCF probabilities of the four redundant check valves for the mixed, staggered, and non-staggered testing schemes. The CCF probabilities of the four redundant check valves for the mixed testing schemes were estimated to be higher than those for the staggered testing scheme, and lower than those for the non-staggered testing scheme.
[en] A thermohydraulic and safety analysis code-TSACC has been developed using Fortran90 language to evaluate the transient thermohydraulic behavior of the China advanced research reactor (CARR) under station blackout accident (SBA). For the development of TSACC, a series of corresponding mathematical and physical models were applied. Point reactor neutron kinetics model was adopted for solving the reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional correlations were supplied. The usual finite difference method was abandoned and the integral technique was adopted to evaluate the temperature field of the plate type fuel elements. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behavior of the CARR. The computational result of TSACC showed the adequacy of the safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of RELAP5/MOD3 and a good agreement was obtained. The adoption of modular programming techniques enables TASCC to be applied to other reactors by easily modifying the corresponding function modules
[en] As part of the evaluation for a severe accident management strategy, a reactor coolant system (RCS) depressurization in optimized power reactor (OPR)1000 has been evaluated by using the SCDAP/RELAP5 computer code. An indirect RCS depressurization by a secondary depressurization by using a feed and bleed operation has been estimated for a small break loss of coolant accident (LOCA) without a safety injection (SI). Also, a direct RCS depressurization by using the safety depressurization system (SDS) has been estimated for the total loss of feed water (LOFW). The SCDAP/RELAP5 results have shown that the secondary feed and bleed operation can depressurize the RCS, but it cannot depressurize the RCS sufficiently enough. For this reason, a greater direct RCS depressurization by using the SDS is necessary for the 1.35 in. break LOCA without SI. A proper RCS depressurization time and capacity leads to a delay in the reactor vessel failure time from 7.5 to 10.7 h. An opening of two SDS valves can depressurize the RCS sufficiently enough and the proper RCS depressurization time and capacity leads to a delay in the reactor vessel failure time of approximately 5 h for the total LOFW. An opening of one SDS valve cannot depressurize the RCS sufficiently enough