scholarly journals Modeling of quality indicators for water demineralization process at power reactors

2021 ◽  
pp. 36-44
Author(s):  
Svitlana Merdukh ◽  
Yuliia Zaporozhets ◽  
Denys Skladannyi ◽  
Serhii Plashykhin
Keyword(s):  
Steam Generator ◽  
Regression Models ◽  
Nuclear Power ◽  
Specific Conductivity ◽  
Hydrogen Index ◽  
Coefficient Of Determination ◽  
Specific Electrical Conductivity ◽  
Electric Pump ◽  
Desalination Plant ◽  
Salt Compartment

Regression models of demineralizing filters for a nuclear power plant will increase the efficiency of the desalination process for turbine condensate. To identify the control object, experimental and statistical research methods were applied. To detect the presence and study the extent of relations between the process factors and output variables, the STATISTICA (StatSoft) analytical system was used. The impact of factors on the studied traits was assessed using a nonlinear estimation module. Upon the analysis of the desalination plant operation, the following initial parameters were identified. These parameters influence the control decisions in managing the object, can be predicted over time and are not random variables such as the values of specific conductivity and hydrogen index. For specific electrical conductivity: Xh is the specific electrical conductivity of the condensate after the desalination plant; Na is the concentration of sodium ions in the condensate behind the condensate electric pump of the first stage; CN2H4 is the concentration of hydrazine in the feedwater of the steam generator; CFe is the concentration of iron ions in the condensate on the condensate electric pump of the first stage; T is the temperature of the condensate sample on the condensate electric pump of the first stage; CCl is the concentration of chlorine ions in the purge water of the salt compartment of the steam generator; and CSO4 is the concentration of sulfate ions in the purge water of the salt compartment of the steam generator. For the hydrogen index: pH is the hydrogen index of the condensate after the desalination plant; CN2H4 is the concentration of hydrazine in the feedwater of the steam generator; T is the temperature of the condensate sample on the condensate electric pump of the first stage; and CFe is the concentration of iron ions in the condensate behind the condensate electric pump of the first stage. To detect the presence and study the extent of relations between the selected factors of relevant process and output variables, a correlation analysis was conducted and input factors were subsequently ranked. The correlation analysis allowed ranking all the variables and their possible interaction effects in a descending degree of impact on the outputs. In accordance with the selected system outputs, based on the ranking of input variables, regression analysis was performed and corresponding mathematical dependences were obtained with alternate inclusion of factors in a decreasing degree of their influence on the output variable to establish functional dependencies between the experimental data on desalination plant operation. The regression dependence for the experimental data of specific conductivity on the selected state parameters was obtained.  The coefficient of determination is equal to 0.7245. The regression dependence describing the change of the hydrogen index has also been obtained. The coefficient of determination is equal to 0.7231. The proposed models explain 72% of the variation in the dependent variable. The correlation ratio is 0.82, which determines the close relationship between the values. Regression models adequately describe the operational data of the nuclear power plant and can be used as part of the control system of the filter unit.

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

Structural Integrity Assessment of Steam Generator Tubes Using a New EPRI Statistical Approach

2002 ◽  
Author(s):  
Carlos Alexandre de Jesus Miranda ◽  
Miguel Mattar Neto
Keyword(s):  
Steam Generator ◽  
Wall Thickness ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Tube Wall ◽  
New Approach ◽  
Inconel 600 ◽  
Integrity Assessment ◽  
Tube Wall Thickness ◽  
Uniform Tube

A fundamental step in tube plugging management of a Steam Generator (SG), in a Nuclear Power Plant (NPP), is the tube structural integrity evaluation. The degradation of SG tubes may be considered one of the most serious problems found in PWRs operation, mainly when the tube material is the Inconel 600. The first repair criterion was based on the degradation mode where a uniform tube wall thickness corrosion thinning occurred. Thus, a requirement of a maximum depth of 40% of the tube wall thickness was imposed for any type of tube damage. A new approach considers different defects arising from different degradation modes, which comes from the in-service inspections (NDE) and how to consider the involved uncertainties. It is based on experimental results, using statistics to consider the involved uncertainties, to assess structural limits of PWR SG tubes. In any case, the obtained results, critical defect dimensions, are within the regulatory limits. In this paper this new approach will be discussed and it will be applied to two cases (two defects) using typical data of SG tubes of one Westinghouse NPP. The obtained results are compared with ‘historical’ approaches and some comments are addressed from the results and their comparison.

The Geometry of Steam Generator Tube and Its Relevance to the Occurrence of Stress Corrosion Cracking in Operating Nuclear Power Plants

2010 ◽  
Author(s):  
Deok Hyun Lee ◽  
Do Haeng Hur ◽  
Myung Sik Choi ◽  
Kyung Mo Kim ◽  
Jung Ho Han ◽  
...  
Keyword(s):  
Stress Corrosion ◽  
Steam Generator ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Corrosion Cracking ◽  
Location Type ◽  
Steam Generator Tubes ◽  
Non Destructive

Occurrences of a stress corrosion cracking in the steam generator tubes of operating nuclear power plants are closely related to the residual stress existing in the local region of a geometric change, that is, expansion transition, u-bend, ding, dent, bulge, etc. Therefore, information on the location, type and quantitative size of a geometric anomaly existing in a tube is a prerequisite to the activity of a non destructive inspection for an alert detection of an earlier crack and the prediction of a further crack evolution [1].

Leak behavior of SCC degraded steam generator tubings of nuclear power plant

2005 ◽  
Vol 235 (23) ◽  
pp. 2477-2484 ◽  
Author(s):  
Seong Sik Hwang ◽  
Hong Pyo Kim ◽  
Joung Soo Kim ◽  
Kenneth E. Kasza ◽  
Jangyul Park ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power

scholarly journals A NON-DESTRUCTIVE METHOD FOR ESTIMATING LEAF AREA OF Ceiba glaziovii (Kuntze) K. Schum.

FLORESTA ◽  
2019 ◽  
Vol 50 (1) ◽  
pp. 1063
Author(s):  
João Everthon da Silva Ribeiro ◽  
Francisco Romário Andrade Figueiredo ◽  
Ester Dos Santos Coêlho ◽  
Walter Esfrain Pereira ◽  
Manoel Bandeira de Albuquerque
Keyword(s):  
Leaf Area ◽  
Regression Models ◽  
Information Criterion ◽  
Coefficient Of Determination ◽  
Linear Measurements ◽  
Maximum Width ◽  
Bias Index ◽  
Different Shapes ◽  
Non Destructive

The determination of leaf area is of fundamental importance in studies involving ecological and ecophysiological aspects of forest species. The objective of this research was to adjust an equation to determine the leaf area of Ceiba glaziovii as a function of linear measurements of leaves. Six hundred healthy leaf limbs were collected in different matrices, with different shapes and sizes, in the Mata do Pau-Ferro State Park, Areia, Paraíba state, Northeast Brazil. The maximum length (L), maximum width (W), product between length and width (L.W), and leaf area of the leaf limbs were calculated. The regression models used to construct equations were: linear, linear without intercept, quadratic, cubic, power and exponential. The criteria for choosing the best equation were based on the coefficient of determination (R²), Akaike information criterion (AIC), root mean square error (RMSE), Willmott concordance index (d) and BIAS index. All the proposed equations satisfactorily estimate the leaf area of C. glaziovii, due to their high determination coefficients (R² ≥ 0.851). The linear model without intercept, using the product between length and width (L.W), presented the best criteria to estimate the leaf area of the species, using the equation 0.4549*LW.

scholarly journals Non-Linear Regression Models with Vibration Amplitude Optimization Algorithms in a Microturbine

Sensors ◽  
2021 ◽  
Vol 22 (1) ◽  
pp. 130
Author(s):  
Omar Rodríguez-Abreo ◽  
Juvenal Rodríguez-Reséndiz ◽  
L. A. Montoya-Santiyanes ◽  
José Manuel Álvarez-Alvarado
Keyword(s):  
Linear Regression ◽  
Vibration Amplitude ◽  
Regression Models ◽  
Engineering Problem ◽  
Coefficient Of Determination ◽  
Model Parameters ◽  
Gray Wolf ◽  
Linear Regression Models ◽  
Machinery Condition Monitoring ◽  
Non Linear

Machinery condition monitoring and failure analysis is an engineering problem to pay attention to among all those being studied. Excessive vibration in a rotating system can damage the system and cannot be ignored. One option to prevent vibrations in a system is through preparation for them with a model. The accuracy of the model depends mainly on the type of model and the fitting that is attained. The non-linear model parameters can be complex to fit. Therefore, artificial intelligence is an option for performing this tuning. Within evolutionary computation, there are many optimization and tuning algorithms, the best known being genetic algorithms, but they contain many specific parameters. That is why algorithms such as the gray wolf optimizer (GWO) are alternatives for this tuning. There is a small number of mechanical applications in which the GWO algorithm has been implemented. Therefore, the GWO algorithm was used to fit non-linear regression models for vibration amplitude measurements in the radial direction in relation to the rotational frequency in a gas microturbine without considering temperature effects. RMSE and R2 were used as evaluation criteria. The results showed good agreement concerning the statistical analysis. The 2nd and 4th-order models, and the Gaussian and sinusoidal models, improved the fit. All models evaluated predicted the data with a high coefficient of determination (85–93%); the RMSE was between 0.19 and 0.22 for the worst proposed model. The proposed methodology can be used to optimize the estimated models with statistical tools.

scholarly journals Aplicación de la Teoría de Perturbación – Método Diferencial- al Análisis de Sensibilidad en Generadores de Vapor de Centrales Nucleares PWR-Caso Angra I

2019 ◽  
pp. 119-126
Keyword(s):  
Perturbation Theory ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Homogeneous Model ◽  
Differential Method ◽  
Differential Approach ◽  
Good Potential ◽  
Direct Calculations

Aplicación de la Teoría de Perturbación – Método Diferencial- al Análisis de Sensibilidad en Generadores de Vapor de Centrales Nucleares PWR-Caso Angra I Aplication of the Perturbation Theory- Differential Methodto Sensibility Análisis in PWR Nuclear Power Plant Steam Generator- Angra I Giol Sanders R, Andrade de Lima F, Marques A, Gallardo A, Bruna M, Zúñiga A Institución Peruano de Energía Nuclear Universidad Federal de Rio De Janeiro-Brasil DOI: https://doi.org/10.33017/RevECIPeru2011.0033/ RESUMEN En este trabajo basado en la tesis del Magíster Roberto Giol S. [1] presenta una aplicación del formalismo diferencial de la teoría de perturbación a un modelo termohidráulico homogéneo de simulación del comportamiento estacionario de uno de los generadores de vapor de la Central Nuclear tipo PWR Angra I del Brasil. Se desarrolla un programa de cálculo PERGEVAP tomando como base el código GEVAP de Souza[2]. El programa PERGEVAP permite realizar cálculos de sensibilidad de funcionales lineales (temperatura media del primario)y no lineales (flujo de calor medio a través de las paredes de los tubos del generador) con relación a las variaciones de ciertos parámetros termo-hidráulicos(flujo másico del primario, calor específico, etc), Los resultados obtenidos con este formalismo son luego comparados con los obtenidos del cálculo directo con el propio código GEVAP, pudiéndose verificar una excelente concordancia. Este método se muestra promisorio para efectuar cálculos repetitivos asociados al diseño y análisis de Seguridad de los componentes de las Centrales Nucleares. Descriptores: teoría de perturbación, método diferencial, sensibilidad, generador de vapor, central nuclear PWR. ABSTRACT This report presents an application of the differential approach of the perturbation theory to an homogeneous model of a PWR steam generator in the Angra 1 Nuclear Power Plan in Brazil under steady-state conditions. Program PERGEVAP was built fom the code GEVAP developed by Souza and allows sensitivity calculations of linear (average primary loop temperature) and non-linear (average heat flux) functionals due to variations in some thermo-hydraulics parameters (flow rate, specific heat, , etc). Results obtained with this approach are then compared with direct calculations performed using the GEVAP code, with excellent agreements. The method has good potential to treat repeated calculations needed in the design and safety analysis of the Nuclear Plant components. Keywords: perturbation theory, differential method, steam generator, PWR nuclear Power Plant.

Sign in / Sign up

Export Citation Format

Share Document