Results 1 - 10 of 117
Results 1 - 10 of 117. Search took: 0.019 seconds
|Sort by: date | relevance|
[en] Highlights: • Wilks’ method for setting tolerance limits is derived and verified. • Higher order Wilks analysis increases the accuracy and precision of the predicted tolerance. • In most practical applications, higher order analysis is unnecessary. • Wilks’ method is applied to the Dittus-Boelter equation. - Abstract: Wilks’ non-parametric method for setting tolerance limits using order statistics has recently become popular in the nuclear industry. The method allows analysts to predict a desired tolerance limit with some confidence that the estimate is conservative. The method is popular because it is simple and fits well into established regulatory frameworks. A critical analysis of the underlying statistics is presented in this work, including a derivation, analytical and statistical verification, and a broad discussion. Possible impacts of the underlying assumptions for application to computational tools are discussed. An in-depth discussion of the order statistic rank used in Wilks’ formula is provided, including when it might be necessary to use a higher rank estimate.
[en] Highlights: • A structural method for uncertainty evaluation of constitutive models is proposed. • Several tools for model uncertainty quantification are investigated and discussed. • Uncertainties of important models during the LBLOCA are quantified. • Best estimate plus uncertainty analysis is applied to the LP-02-6 experiment. - Abstract: Best estimate (BE) codes are developed to carry out realistic safety analysis of the nuclear reactor, and generally hundreds of constitutive models are comprised in a BE code. Nevertheless, uncertainties of these constitutive models are often not properly handled in the best estimate plus uncertainty (BEPU) analysis. It is not sufficient to fully evaluate the uncertainties of different sorts of models with only one method. Thereby a structural method for uncertainty quantification (UQ) of constitutive models is proposed and relevant description is presented. Based on the method, constitutive models will be classified into two categories according to the characteristics, namely the independent model and the dependent model, and different methods will be adopted for different sorts of model. Several statistical methods for UQ of constitutive models such as the non-parametric curve estimation method, the Bayesian calibration method and the coverage calibration method are evaluated and the characteristics of these methods are discussed. In addition, several methods for the construction of surrogate model are utilized to reduce the computational cost, and a model selection technique is adopted to opt the optimal model among all alternative models. The large break loss of coolant accident (LBLOCA) experiment LOFT LP-02-6 is utilized to verify the proposed structural method, and the BEPU analysis of the LP-02-6 experiment is carried out. The results show that uncertainty intervals of the identified models obtained through the structural method are reasonable, uncertainties of the peak cladding temperature (PCT) as well as the accumulator injection time (AIT) are quantified, and the sensitivity analysis is carried out to evaluate the influence of different input parameters on the 1st PCT, 2nd PCT and the AIT.
[en] Highlights: • Safety critical systems are designed to function in safe manner so that its failure should not lead to the catastrophic effects. • Due to safety significance of such systems, these have high performance requirements. • The strategy discussed for performance analysis of safety critical and control systems and to estimate performance based risk factor. • The technique elaborates Petri nets to estimate performability to ensure system dependability requirements. • The technique has been validated on 17 safety critical and control systems of Nuclear Power Plant. - Abstract: Non-functional requirements play a critical role in designing variety of applications domain ranging from safety-critical systems to simple gaming applications. Performance is one of the crucial non-functional requirements, especially in control and safety systems, which validates the design. System risk can be quantified as a product of probability of system failure and severity of its impact. In this paper, we devise a technique to do the performance analysis of safety critical and control systems that helps to estimate the risk. The technique elaborates Petri nets to estimate performability to ensure system dependability requirements. We illustrate the technique on a case study of Nuclear Power Plant. The technique has been validated on its 17 safety critical and control systems.
[en] The reactor protection system (RPS) in a research reactor is a well-known conventional setpoint-based protection system. The RPS performs protective actions with the generation of alarms when the measurement values exceed the setpoints. The RPS has disadvantages in that alarms are not generated before the measurement values exceed the setpoints; they are generated at the time of protection actions are performed. In addition, each alarm has a direct relation with signals, not accidents, so it is difficult to identify the accident type quickly. Thus, new methods are required to diagnose and classify accidents. We propose a deep-learning-based alarm system. The proposed alarm system is modeled with convolutional and fully connected neural networks. The proposed scheme is designed from safety analysis in the safety analysis report. We prepare various datasets and scenarios for training and test. The results show that the proposed alarm system provides fast diagnosis alarms with probability values.
[en] Highlights: • The characteristics of density wave oscillation in the natural circulation system were analyzed. • The effects of heaving and rolling motions on flow instability in the natural circulation system were analyzed. • The resonance phenomenon between density wave oscillation and ocean motions was found and analyzed. - Abstract: The study of natural circulation is attracting more and more attention of researchers and engineers for the improvement of the system safety. A lot of natural circulation systems are applied to new generation reactors which make reactor systems safer and more compact because there are fewer components and no circulation pump in the natural circulation systems. The residual heat can be removed passively from the reactor by the natural circulation system which prolongs the response time for operators and reduces the potential accident because of human error. With the application of nuclear reactors under the ocean conditions, the study on flow oscillation characteristics in natural circulation system under ocean condition becomes very significant. Usually, flow oscillation is harmful to the operation of natural circulation systems which can affect the cooling capability of natural circulation systems in some extreme situations. In this paper, the classification of flow instability is introduced for systems under ocean conditions. Characteristics of density wave oscillations (DWO) under natural circulation are studied, including oscillation characteristics, stability boundary and subcooling effect. The characteristics of DWO under rolling and heaving motions are compared with those under the normal condition. The effects of rolling and heaving characteristics are analyzed, including maximum rolling angle, rolling period, rolling center location, maximum acceleration and heaving period. The resonance phenomenon is found, and resonant behaviors are analyzed.
[en] Highlights: • A validated numerical UTSG model on CFD method is established and studied. • The influence of total mass flow rate of UTSG on inversion in UTSG is studied. • To do mechanism analysis of reversion, the impact of inlet initial flow rate of U-tube on reversion is studied. • It is more favorable to reduce reversion while increasing total inlet mass flow rate of UTSG. • The U tube with small inlet initial mass flow rate is more likely to happen reverse flow. - Abstract: In view of the phenomenon of reverse flow in partial inverted U-tubes of inverted U-tube steam generator (UTSG) under natural circulation condition, the numerical calculation method is used to study the flow and heat transfer in the primary side of UTSG. There are many factors influencing reverse flow in UTSG, such as the inverted U-tube length, coolant mass flow, operating pressure and temperature difference between primary and secondary side. However, in a specific UTSG, the lengths of different inverted U-tubes and the variations of internal mass flow rates are not uniform, so it is difficult to study the reverse flow mechanism. In this paper, we study the reverse flow of UTSG with identical tubes. Results show that under natural circulation condition, it is more favorable to reduce reverse flow as total inlet mass flow rate in UTSG increases. In addition, the critical value of reversion is calculated, which provides a reference for the engineering design and reactor safety analysis of UTSG. The spatial distribution of the reverse tubes is related to the inlet initial mass flow rate of U-tube and the U-tube with small inlet initial mass flow rate is more likely to happen reverse flow. U-tube with least inlet mass flow rate firstly triggers to the inflexion point of reversion, and the second least one happens reverse flow next while the total inlet mass flow rate in UTSG decreases gradually.
[en] Highlights: • Effects of temperature and pressure variations on gas accumulation is investigated. • Pressure change in a closed system does not affect the gas accumulation. • The higher the water temperature, the more gases are accumulated. • More than 50% of accumulated gas remains, when the water is cooled back. • Gas is accumulated if solubility of dissolved gas is lower than initial condition. - Abstract: Unexpected gas accumulation in a safety system of nuclear power plants can damage system elements and degrade cooling performance. To prevent the gas accumulation, it is important to define the mechanism of gas accumulation occurred by separation of dissolved gas. In the current research, it is investigated how the variations of temperature and pressure affect separation of dissolved gases and accumulation of non-condensable gases. The experimental study is conducted on three subjects; gas accumulations by variations of (1) pressure, (2) temperature and (3) order of temperature and pressure changes. Before performing each experiment, demineralized water is stabilized under the initial condition (20 °C and atmospheric pressure) for more than 24 h. In a closed system, the gas accumulation cannot be occurred by pressure change. By heating water, the gas accumulation is generated. When the water is kept in higher temperature, the more gases are accumulated. Even if the water is cooled back to the initial condition, the accumulated gas is remained more than 50%. By changing the order of temperature and pressure variations, the gas is accumulated. In results, it is found that the gas accumulation can be generated if the solubility of dissolved gas becomes lower than initial condition. Additionally, it is found that the gas already accumulated in the system is difficult to remove without additional venting process.
[en] Highlights: • The scaling analysis for the critical transition period in a SBLOCA was conducted. • A uniform scaling law for existing T-branch entrainment models was obtained. • The requirement for reproducing the synergetic effect in an IET was introduced. • The transition behavior were investigated on integral effect test. • The system configuration influences on the core safety were studied. - Abstract: The passive safety commercial pressurized water reactors (PWRs), AP1000 and CAP1400 are equipped with the automatic depressurization system (ADS). The fourth stage of ADS (ADS-4) has the largest discharging capacity among the all ADS stages. During a small break loss-of-coolant accident (SBLOCA), only when the system is fully depressurized by the ADS-4 to near the containment atmospheric pressure, the in-containment refueling water storage tank (IRWST) can deliver the large coolant inventory for the long-term core cooling by its water head. Before the IRWST injection, the core reaches its minimum inventory. So the transition from the ADS-4 depressurization to the IRWST injection is a very challenging period in a SBLOCA. Therefore, to evaluate the passive system performance by the integral effect test (IET), the transition must be simulated properly. This work conducted the scaling analysis for the ADS-4 depressurization and the IRWST injection for designing the related components of the ACME IET facility. A uniform scaling law based on the existing T-branch liquid entrainment models was proposed. For scaling the onset of the IRWST injection, the requirement for reproducing the synergetic effect in an IET was introduced in order to preserve the important event/process occurrence time with the proper conditions, and the scaling laws for the full pressure IRWST injection were obtained. For investigating the transition thermal hydraulic behaviors and the ADS-4 configuration influences on the core safety, one cold leg break test and three double-ended direct-vessel-injection (DEDVI) line break tests with the different ADS-4 failure modes were selected from the ACME SBLOCA test matrix, and the main test data related to the core safety were analyzed and compared. It was found that the ADS-4 actuation significantly accelerated the system depressurization but worsened the core inventory depletion at the same time. In addition, the break condition and the ADS-4 failure mode could differently affect the system depressurization and the core level. The ACC injection during the ADS-4 depressurization in DEDVI was critical for the core level makeup. The insufficient ADS-4 venting capacity delayed IRWST injection, but also mitigated the core inventory depletion. It suggests the ADS-4 design can be potentially optimized in the passive safety system design.
[en] Highlights: • A thermal hydraulic transient analysis code for OFNPs is developed. • Ocean condition models are established by considering the effects of ship motions. • The developed code is verified by experimental data under rolling motion. • Effects of the coupled ship motion on natural circulation system are studied. - Abstract: Offshore floating nuclear power plants (OFNPs) can effectively solve the offshore energy supply problem in marine resource development and island construction. Affected by ocean waves and other ocean conditions, the OFNPs can generate different kinds of ship motions, which can oscillate the thermal hydraulic parameters and threaten the reactor safety. In the present study, ocean condition theoretical models are established by considering the effects of three basic movement forms (static inclining, linear motion and rotation motion) as well as the coupled ship motions. A thermal hydraulic transient analysis code for OFNPs is developed by adding ocean condition theoretical models into the RELAP5/SCDAPSIM/MOD3.4 code. The experimental data obtained by zero power loading experiment and single-phase natural circulation experiment under rolling motion are used to verify the ocean condition theoretical models as well as the code modification strategy. Results show that the flow fluctuation behaviors caused by rolling motion can be well simulated by the developed code. The calculation capability of modified RELAP5 code under static inclining and heaving motion is also verified by comparing with RETRAN-02/GRAV code. Besides, the effects of the coupled ship motions on natural circulation system are studied by the modified RELAP5 code. Compared with the basic movement forms, the coupled ship motions can cause greater flow fluctuation and obviously reduce the core flow rate, which means the influence of the coupled ship motions is necessary to be considered in the safety analysis of OFNPs.
[en] Highlights: • A Polar Bear Optimization (PBO) algorithm is employed for core loading pattern optimization problem. • A comprehensive optimization package using PARCS and COBRA-EN code with PBO has been developed. • Important neutronic and thermal-hydraulic safety parameters are optimized during the first cycle of BNPP reactor. • The optimized pattern that found with PBLPO improved safety parameters comparing to loading pattern that the proposed in the Final Safety Analysis Report. - Abstract: Loading Pattern Optimization (LPO) is an extremely important issue for the safe and economic operation at nuclear power plants because the core configuration affects many of the neutronic and thermal-hydraulic parameters (NTPs) of the reactor. Most researches were done in this area limited to employing different heuristic optimizing algorithm and the neutronic aspects at the beginning of the cycle (BOC). In this research, a comprehensive code package has been developed using coupling the PARCS code and the COBRA-EN code with the new Polar Bear Optimization (PBO) algorithm, which tries to find an optimal loading pattern with maximized cycle length, maximized departure from the nucleate boiling ratio (DNBR), and flatter power distribution. Other core safety parameters such as maximum boric acid concentration at the BOC, maximum power peaking factor (PPF), the maximum temperature of the fuel center and the maximum clad surface temperature are considered as penalties for accepting or rejecting optimal configuration alongside the fitness function. This code Package which named PBLPO (Polar Bear Loading Pattern Optimization) code, was implemented for the core of VVER-1000 reactor in Bushehr Nuclear Power Plant (BNPP). The optimized pattern that found with PBLPO improved safety parameters comparing to loading pattern that the proposed in the Final Safety Analysis Report (FSAR) and shows that the polar bear algorithm has a good performance in LPO.