Bushehr Nuclear Power Plant Research Articles

Fire simulation in the main control room of Bushehr nuclear power plant using CFAST software

Fire simulation in the main control room of Bushehr nuclear power plant using CFAST software

Compilation of the Job Competencies Model of the Parent Company Specialized in the Production and Development of Atomic Energy in Iran

Objective: The aim of this study is to develop a competency model for jobs at the Bushehr Nuclear Power Plant, a subsidiary of the Atomic Energy Organization of Iran. Methodology: This mixed-methods research utilized both qualitative and quantitative approaches. Initially, qualitative data were collected through library studies and thematic analysis. Based on this data, a questionnaire was designed to gather quantitative information, which was distributed among 90 employees and managers of the Bushehr Nuclear Power Plant. The collected data were then analyzed using structural equation modeling. Findings: The results of the analysis showed that the proposed competency model had a good fit, with all structural paths being statistically significant. Both technical and managerial competencies played a crucial role in improving employee performance and safety. The model’s goodness-of-fit index was calculated at 0.413, indicating a strong fit. Conclusion: This study concluded that developing a multidimensional competency model, including technical, managerial, and communication skills, can lead to improved efficiency and safety in nuclear power plants. Implementing such models in recruitment, training, and employee evaluation processes can positively impact organizational productivity.

Economic evaluation of hydrogen production in second and third units of Bushehr nuclear power plant regarding future need of nuclear fission technology

Today, the most important challenge for hydrogen production is the use of suitable energy source for sustainable production. The dual use of nuclear power plants allows them to expand into other markets by eliminating greenhouse gas emissions from processes such as hydrogen production. The significance of this matter especially in countries with ongoing new nuclear power plants is high and shows the future need for nuclear fission technology. In this paper, an economic assessment of hydrogen production in the second and third units of the Bushehr nuclear power plant (BNPP) with the Hydrogen Economy Evaluation Program (HEEP) is carried out and a comparison of CO2 emission in nuclear and non-nuclear methods of hydrogen production is presented. Most nuclear reactors, including Advanced Pressurized WaterReactor (APWR), High Temperature Gas Reactor (HTGR), and Water-Water EnergeticReactor (WWER), can be used as a source of thermal and electrical energy for the process of hydrogen production. In this research, hydrogen production processes based on the Conventional Electrolysis (CE), High Temperature Steam Electrolysis (HTSE), and Sulfur-Iodine (SI) Process using different storage and transportation options are compared to choose the cost-effective method for coupling with the WWER1000 reactor. Hydrogen production, thermal energy, and electricity costs have been calculated and explained in detail. The results showed that the cost-effective option for hydrogen production is using the High Temperature Steam Electrolysis (HTSE) method, Compressed Gas (CG) storage, and pipeline transportation, by 2.88 $/kg. Also, applying this method to produce hydrogen, the nuclear power plant thermal efficiency, and the hydrogen plant efficiency are kept in the best possible conditions by 51.98% and 58.21%, respectively.

The large-break LOCA uncertainty analysis in a VVER-1000 reactor using TRACE and DAKOTA

An uncertainty analysis is inevitable in simulating reactor performance, particularly in the accident analysis for the estimation of safety borders. In this paper, the application of the uncertainty analysis of a Large Break Loss of Coolant Accident (LBLOCA) from main coolant pipeline at the reactor inlet and outlet, and also pressurizer (PRZ) connecting pipe are presented. First, the model developed in TRACE is verified in steady-state conditions and LBLOCA using the Final Safety Analysis Report (FSAR) of Bushehr nuclear power plant. By the verified model, the uncertainty analysis is made respecting uncertainty in boundary conditions. Finally, the analysis is achieved using the surrogate models and Wilks’ method in DAKOTA. The 95/95 uncertainty band of the peak cladding temperature (PCT) and main reactor parameters are obtained and analyzed. The results indicate that the PCT uncertainty in LBLOCA at the reactor inlet is large, the PCT varies in the range of 910 to 970 °C. The drop temperature time of fuel cladding is in the range of 270 to 600 s for LBLOCA at the reactor inlet. The PCT uncertainty in LBLOCA at the reactor outlet and PRZ connecting pipe is low.

Investigation of the source term and radiological consequences of small break LOCA with failure of the emergency core cooling system in Bushehr nuclear power plant

In the present study, the in-containment source term derivation, environmental dispersion and deposition as well as radiological consequences of Bushehr Nuclear Power Plant (BNPP) due to Small Break Loss of Coolant Accident (SB-LOCA) have been studied. First, the calculation of in-containment release and distribution of the source terms due to the severe accident of SB-LOCA with failure of the active part of the Emergency Core Cooling System (ECCS) has been carried out. In this calculation, two scenarios, including with and without operator control actions, have been considered. Then, the environmental leakage of radionuclides as well as their dispersion and deposition considering two meteorological scenarios have been investigated. Consequently, public exposure has been calculated for each meteorological scenario and the obtained results have been reported and discussed. For two accidents with the same meteorological conditions, at a distance of 400 m from the release point, the received dose in the first 24 h is 18 and 342 mSv without and with operator action, respectively. The results show that due to the intervention of the operator and the use of emergency cooling supply systems, more water enters the reactor. Therefore, if the operator's actions do not prevent the core from melting, more source terms have entered the containment after the core melts and worse conditions than the baseline scenario without operator action occurred.

Advanced fuel and burnable absorbers designed for long-cycle operation of BNPP

The present study is based on the facts that are happening for the Bushehr Nuclear Power Plant (BNPP) and coincides with the operational changes and step-by-step modifications of the BNPP in the near future. Bushehr reactor core will change from BNPP core to TVS-2M core. In other words, a new fuel, called the TVS-2M, is being used by the Russians to upgrade WWER-1000 reactors to reduce power generation costs, increase cycle length and fuel consumption. The neutronic analysis performed in this work is in line with this change. It provides a feasibility study of BA change and increase of cycle length in BNPP reactor without changing the initial structure of the reactor core. This research provides a new core configuration that can achieve the stated goals to a large extent without changing the BNPP core to the TVS-2M core. This research provides a feasibility study of Integrated burnable absorber (IBA) change by maintaining the initial neutronic conditions and a new high-performance configuration aiming for a long-cycle operation in Iran's nuclear power reactor. In Water-Water Energetic Reactors (WWER) reactors, two alternatives to CrB2Al (current burnable absorber in BNPP) are gadolinia (Gd2O3-UO2), and erbia (Er2O3-UO2). The benefits of the combinational IBA concept are expected by compensating weak points of each other. The main contribution of this research is the use of two cylindrical coaxial IBA layers with the same volume of gadolinium pins plus erbium (separated by a thin layer of zirconium alloy (Zr + 1% Nb)) and gadolinium pins. A neutronic evaluation has been conducted to investigate the variations of effective multiplication factor versus burnup, reactivity swing, power peaking distribution, moderator temperature coefficient (MTC), critical boron concentration (CBC), and cycle length. The result shows that the cycle length can be extended from 296 days in the BNPP core to 366 days in the proposed core, which is comparable to the cycle length of the TVS-2M core. The average BA concentration in the proposed core is 5%, which is lower than the average BA concentration of 5%–8% in the TVS-2M core. Regarding the gain of 73.90 days in cycle length, the revenue is estimated to be around 19,503,000 USD, making it worth the implementation considering the number of gadolinium and erbium pins in the proposed core.

Investigation of small break LOCA with failure of the emergency core cooling system accident management in VVER-1000 reactor type

In this study, the thermohydraulic condition of the power plant is estimated as a result of the Small Break Loss of Coolant Accident (SB-LOCA) for Bushehr Nuclear Power Plant (BNPP) and the effect of control measures in postponing severe events as well as increasing the time available to enter control measures has been investigated. Two scenarios including control measures with and without operator action have been considered, considering the progress of the accident and the lack of effective intervention of control measures, both scenarios lead to the melting of the reactor core. The critical times for entering additional measures and restoring systems has been obtained from the progress of the accident for two scenarios that is used to develop an accident management strategy. The thermohydraulic calculation in this article deals with the estimation of time of critical point for accident management measures at a different phase of SB-LOCA.

Evaluation of different integrated burnable absorber materials in fuel assemblies of Bushehr WWER-1000 nuclear reactor

Abstract Nowadays, reducing fuel consumption, fuel cost, and fuel waste is obtained by increasing the burnup and reactor cycle length as essential goals in nuclear power plants. Due to the scarcity of Iran’s uranium resources and the plan for using new types of fuel in the Water–Water Energetic Reactors (WWER-1000) at Bushehr Nuclear Power Plant (BNPP), utilizing Integral Burnable Absorbers (IBAs) is of great importance. In the present study, a neutronic evaluation has been conducted to investigate the variations of infinite multiplication factor versus burnup, reactivity swing, and power distribution in various IBAs, including Gd2O3-UO2, Er2O3-UO2, and Dy2O3-UO2. The results were compared with the standard burnable absorber in the BNPP reactor core (CrB2Al). It can be concluded that gadolinium IBA, with a concentration of 5%, has the greatest effect on the initial reactivity and reduces the reactivity swing by 19% compared to the CrB2Al BA. In addition, by using coaxial gadolinium-erbium IBA pins, as well as optimizing the neutronic condition, the reactor cycle length increased by 1.01 GWd/MTU compared to the standard Bushehr BA.

Study and assessment of the newly proposed passive management of MSLB + SBO accidents for the VVER 1000/V446 NPP using the secondary side Deaerator tank

To manage accidents in actual conditions, we need to evaluate different accident scenarios on nuclear power plants. Removing the residual heat from the reactor core after the loss of off-site power (LOOP) accident and Station black-out (SBO) accident is one of the biggest challenges in such an incident. In this paper, LOOP and SBO Accidents with the main steam line break (MSLB) accident of No. 2 (inside the Containment) have been simulated by the thermal-hydraulic RELAP5 code on the VVER-1000/v446 reactor. A total of 5 different scenarios have been simulated, assuming the most difficult conservative conditions, and the results reveal the critical role of diesel generators in supplying emergency feed water pumps (EFWP), as well as the part of operator performance in securing a nuclear power plant in the condition of an accident. In the case of using a series of operator operations and using the water resources of the Deaerator tank on the secondary side with the Feed and Bleed system on the primary side, in The most brutal conservative mode (MSLB + SBO accident) by injecting the water into the steam generator-4, the temperature and pressure of the coolant in the primary side are reduced and by absorbing more heat from the reactor, core melting is delayed about 25,000 s, and the safety of Bushehr nuclear power plant (BNPP) against this accident will be improved.

PCTRAN enhancement for large break loss of coolant accident concurrent with loss of offsite power in VVER-1000 simulation

Abstract PCTRAN capability to simulate a large break loss of coolant accident concurrent with the loss of offsite power in Bushehr Nuclear Power Plant is enhanced and investigated. Following the correction of the accident scenario for Bushehr nuclear power plant in PCTRAN, simulation results are compared with the final safety assessment report of that plant. As a result, the primary loop thermal hydraulics parameters including pressure, total flow rates, leakage flow rates and reactor power are in a good agreement with the reference data. Hot and cold leg temperature variations have the same trends as reference data but have a maximum of 80 8C disagreement at the transient initiation. The reason for this disagreement is explained and its adjustment is discussed. Improvements of PCTRAN simulator are mainly due to enhancing user control for atmospheric steam dump valve, containment pressure and emergency core cooling systems which are thoroughly described in this paper.

Evaluation of advanced accumulator in a VVER-1000 reactor in loss of coolant accident

The advanced accumulator is a passive system which gradually decreases the high flow into a low flow. The role of the fluidic device is to obtain an additional safety margin by increasing golden time to cope in the LOCA. In this paper, the effect of using the safety injection tank and the fluidic device, called advanced accumulator, in a wide spectrum of LOCA break sizes without emergency core cooling systems (ECCS) of a VVER-1000 is evaluated. Firstly, the main component of primary and secondary loops of Bushehr nuclear power plant (BNPP) is modeled by RELAP5. Secondly, the performance of the model is assessed in steady-state and LOCA conditions by comparing the obtained results with the plant’s FSAR. Thirdly, a model is developed in RELAP5 to simulate the behavior of advanced accumulator and validated by blowdown test in a test facility and MARS-KS code. Finally, the performance of advanced accumulator in the spectrum of 200–300 mm break sizes of LOCA without ECCS. In this range of breaks, the core damage times (CDT) developed from peak cladding temperature (PCT) of 1204 °C have been calculated. In accordance with the results, the advanced accumulators can increase the CDT in the whole range of breaks.

Investigation of the Source Term and Radiological Consequences of Small Break LOCA with Failure of the Emergency Core Cooling System in Bushehr Nuclear Power Plant

Investigation of the Source Term and Radiological Consequences of Small Break LOCA with Failure of the Emergency Core Cooling System in Bushehr Nuclear Power Plant

ტექნოლოგიური პროცესების მართვის ავტომატიზებული სისტემების უსაფრთხოება, SCADA – შეტევის ობიექტი

(……) Modern automated technology process control systems and the chances of attacks on them are examined in this article. It studies worm virus, Stuxnet, and its detection at the Bushehr Nuclear Power Plant. It also analyzes ways of carrying out attacks on critically important objects, and provides analytical tools for the security of technological process systems. The ways for discovering nodes compromised by the Stuxnet virus are proposed. The article considers technological network of typical topology and its typical vulnerabilities; it analyzes the Modbus protocol, the routing system, and passwords on Cisco routers.

Radioactive impact on Iran and the world from a postulated accident at Bushehr Nuclear Power Plant

Chernobyl and Fukushima nuclear accidents showed that nuclear accident is a global concern. Dispersion and faraway transportation of radioactive materials under atmospheric conditions are the main goals of this paper. Radioactive materials have a lot of damages, undesirable and dangerous long-term effects on the environment and human health. Therefore, in this paper, the simulation of transport, dispersion and deposition of radioactive materials released from a postulated accident in Bushehr Nuclear Power Plant have been studied using HYSPLIT. The postulated accident for Bushehr Nuclear Power Plant is designed based on world's largest nuclear accident, the Chernobyl accident. In this study, it is assumed that the postulated accident in Bushehr nuclear plant occurred on March 21, 2015. The simulation results show that in the period of one month, nuclear pollution resulting from this accident would affects almost the whole of Iran. The amount of radioactive materials deposition is very high in some cities of Iran. The most deposition of radioactive materials would occur in Bandar Bushehr and Ahvaz. In these cities, the maximum deposition is related to the second day of the accident, which are: 1.55×108Bq/m2 and 1.06×106Bq/m2, respectively. During this period, in addition to Iran, even most Asian countries, North America, Europe continent and Africa are affected by nuclear pollution while the most pollution is related to Asia, North America and Europe, respectively. Radioactive materials after passing from the continents come back to the East Asian again. In Asian countries after Iran, the maximum deposition of radioactive materials is related to Turkmenistan that the deposition value is 1.78×105Bq/m2 for the second day of the accident. In continent of America, the most nuclear pollution is related to the state of Arizona. In this state, the maximum deposition is 1.00×105Bq/m2 on the fifteenth day of the accident. In Europe, the maximum deposition is in Ukraine and it is 4.16×103Bq/m2 on 12 days after the accident. In this accident, radioactive materials would contaminate the most of Northern hemisphere and also would influence the southern hemisphere in very negligible amount.

Detection and estimation of faulty sensors in NPPs based on thermal-hydraulic simulation and feed-forward neural network

Sensors are one of the most vital instruments in Nuclear Power Plants (NPPs), and operators and safety systems monitor and analyze various parameters reported by them. Failure to detect sensors malfunctions or anomalies would lead to the considerable consequences. In this research, a new method based on thermal–hydraulic simulation by RELAP5 code and Feed-Forward Neural Networks (FFNN) is introduced to detect faulty sensors and estimate their correct value. For design an efficient neural net, seven feature selectors (i.e., Information gain, ReliefF, F-regression, mRMR, Plus-L Minus-R, GA, and PSO), three sigmoid activation functions (i.e., Logistic, Tanh and Elliot), and three training algorithms (i.e., Levenberg–Marquardt (LM), Bayesian Regularization (BR) and Scaled Conjugate Gradient (SCG)) have been comprehensively compared and evaluated. The required data have been obtained by simulating LOFA and SBLOCA transients in RELAP5 code for the Bushehr Nuclear Power Plant (BNPP). The main advantage of this method is that with the failure of more than one sensor, the detection of other sensors is not completely disrupted, and are monitored continually and independently.

The ensemble approach in comparison with the diverse feature selection techniques for estimating NPPs parameters using the different learning algorithms of the feed-forward neural network

Several reasons such as no free lunch theorem indicate that there is not a universal Feature selection (FS) technique that outperforms other ones. Moreover, some approaches such as using synthetic dataset, in presence of large number of FS techniques, are very tedious and time consuming task. In this study to tackle the issue of dependency of estimation accuracy on the selected FS technique, a methodology based on the heterogeneous ensemble is proposed. The performance of the major learning algorithms of neural network (i.e. the FFNN-BR, the FFNN-LM) in combination with the diverse FS techniques (i.e. the NCA, the F-test, the Kendall's tau, the Pearson, the Spearman, and the Relief) and different combination techniques of the heterogeneous ensemble (i.e. the Min, the Median, the Arithmetic mean, and the Geometric mean) are considered. The target parameters/transients of Bushehr nuclear power plant (BNPP) are examined as the case study. The results show that the Min combination technique gives the more accurate estimation. Therefore, if the number of FS techniques is m and the number of learning algorithms is n, by the heterogeneous ensemble, the search space for acceptable estimation of the target parameters may be reduced from n × m to n × 1. The proposed methodology gives a simple and practical approach for more reliable and more accurate estimation of the target parameters compared to the methods such as the use of synthetic dataset or trial and error methods.

Investigating the performance of the supervised learning algorithms for estimating NPPs parameters in combination with the different feature selection techniques

Several reasons such as no free lunch theorem indicates that any learning algorithm in combination with a specific feature selection (FS) technique may give more accurate estimation than other learning algorithms. Therefore, there is not a universal approach that outperforms other algorithms. Moreover, due to the large number of FS techniques, some recommended solutions such as using synthetic dataset or combining different FS techniques are very tedious and time consuming. In this study to tackle the issue of more accurate estimation of NPPs parameters, the performance of the major supervised learning algorithms in combination with the different FS techniques which are appropriate for parameters estimation is considered. The target parameters/transients of the Bushehr nuclear power plant (BNPP) are examined as the case study. By comparing three major supervised learning algorithms (i.e. the MLP-BR, the MLP-LM, and the SVM) in combination with six principal FS techniques (i.e. the NCA, the F-test, the Kendall’s tau, the Pearson, the Spearman, and the Relief) for estimation of three important parameters of NPP (i.e. FMT, CMT, and the DNBR), the BR learning algorithm gives the more accurate results. Therefore, the results show that if the number of FS techniques is m and the number of learning algorithms is n, the search space for more accurate estimation of the NPPs important parameters can be reduced from n × m to 1 × m.

Evaluation of management measures on a hypothetical accident in Unit-1 of bushehr nuclear power plant

The evaluation of various measures on the thermohydraulic conditions of the circuit and especially the fuel conditions of the reactor in order to minimize the accident effects is vital. Therefore, this work examines the effect of the operator's actions on controlling loss of heat removal through the turbine condenser accident with failure of feedwater supply to steam generators.The BNPP parameters during this accident has been analyzed considering two scenarios (with and without operator action) using the RELAP5 code. The simulation results showed that in case 1, the fuel cladding temperature has risen to 1200 °C within 11,000 s, which is more than the allowed temperature. The operator's actions in case 2 maintain the fuel cladding temperature in the permissible range level. Moreover, these results determine the available time for the operator's actions to prevent further core damage under the conditions that have been considered in the selected scenarios.

Identification of the appropriate architecture of multilayer feed-forward neural network for estimation of NPPs parameters using the GA in combination with the LM and the BR learning algorithms

In this study, accurate estimation of nuclear power plant (NPP) parameters is done using the new and simple technique. The proposed technique using the genetic algorithm (GA) in combination with the Bayesian regularization (BR) and Levenberg- Marquardt (LM) learning algorithms identifies the appropriate architecture for estimation of the target parameters. In the first step, the input patterns features are selected using the features selection (FS) technique. In the second step, the appropriate number of hidden neurons and hidden layers are investigated to provide a more efficient initial population of the architectures. In the third step, the estimation of the target parameter is done using different architectures of multilayer feed-forward neural network (MFFN) with LM learning algorithm in which the maximum number of hidden neurons and the maximum number of hidden layers have been limited. In the fourth step, the proposed technique using the GA in combination with the BR learning algorithm is proposed to determine the more appropriate number and the more appropriate distribution of hidden neurons. To study the performance of the proposed technique, Bushehr nuclear power plant (BNPP) transients are examined. The different important transients/parameters are estimated. The results of the estimations by the identified architecture in comparison with the other appropriate architectures show superiority of the proposed technique. Therefore, the proposed technique can be used reliably for accurate estimation of the important parameters and can be used as a support tool by the operators in confront with transients.

VVER-1000/V446 spent fuel pool risk assessment and support through portable mitigating equipment

The Spent Fuel Pool (SFP) accident at the Fukushima Daiichi unit 4 demonstrated the need for identifying and analyzing possible accident sequences and subsequent application of severe accident mitigating systems in LWR (Light Water Reactor) SFPs. In this study, Level 1 Probabilistic Safety Assessment (PSA) is applied to evaluate the Fuel Damage Frequency (FDF) in SFP of Bushehr Nuclear Power Plant unit-1 (BNPP-1). This paper specifically focuses on assessing the SFP risk using Portable Mitigating Equipment (PME) including a reliable robust Portable Diesel Generator (PDG), a Portable Diesel Pump (PDP), and an SFP-spray system proposed to maintain steady decay heat removal. PME is the recommended equipment proposed by IAEA -following the Fukushima disaster- to mitigate accident consequences. In this study, the fault/event trees have been applied to conduct PSA using the SAPHIRE code. The results indicate that the most dominant Initiating Events (IEs) in the reference plant are the Loss of Offsite Power (LOOP) and Loss of Service Water (LOSW), both simultaneously affecting the reactor core and the SFP. There is a significant decrease in calculated FDF – after implementing PDG and PDP- in respect of the reference plant from 3.086E to 6 to 2.166E−7 in the normal storage mode. Heavy load drop, which is a low probability/high consequence IE, becomes crucially important in cases where PME is employed as it directly leads to fuel damage. It is expected that the analyses set forth here would provide efficient guidelines and insights into the strengths and weaknesses inherent in the design and operation of SFP in VVER-1000/V-446 nuclear power plant.