ENCIT2012-090 NEUTRON \& THERMO - HYDRAULIC MODEL OF A REACTIVITY TRANSIENT IN A NUCLEAR POWER PLANT FUEL ELEMENT - PDF
José de Jesús Rivero Oliva, UFRJ, Brazil
Abstract: A reactivity transient without reactor scram was modeled and calculated using analytical expressions for the space distributions of the temperature fields, combined with discrete numerical calculations for the time dependences of thermal power and temperatures. The transient analysis covered the time dependencies of reactivity, global thermal power, fuel heat flux and temperatures in fuel, cladding and cooling water. The model was implemented in Microsoft Office Excel, dividing the Excel file in several separated worksheets for input data, initial steady-state calculations, calculation of parameters non-depending on eigenvalues, eigenvalues determination, calculation of parameters depending on eigenvalues, transient calculation and graphical representation of intermediate and final results. The results show how the thermal power reaches a new equilibrium state due to the negative reactivity feedback derived from the fuel temperature increment. Nevertheless, the reactor mean power increases 40\% during the first second and, in the hottest channel, the maximum fuel temperature goes to a significantly high value, slightly above 2100 0C, after 8 seconds of transient. Consequently, the results confirm that certain degree of fuel damage could be expected in case of a reactor scram failure. Once the basic model has being established the scope of accidents for future analyses can be extended, modifying the nuclear power behavior (reactivity) during transient and the boundary conditions for coolant temperature. A more complex model is underway for an annular fuel element.
Keywords:Non-stationary heat transfer; analytical method; nuclear fuel rod; temperature field; reactivity transient.
Presentation Schedule: Thursday, 10:00-10:20. Session: NCL-2. Presenter: José de Jesús Rivero Oliva.
ENCIT2012-321 COMPUTATIONAL SIMULATION OF TURBULENT NATURAL CONVECTION IN A VOLUMETRICALLY HEATED SQUARE CAVITY - PDF
Camila Braga Vieira, UFRJ, Brazil
Bojan Niceno, Paul Scherrer Institute, Switzerland
Su Jian, UFRJ, Brazil
Abstract: This work aims to analyze the turbulent natural convection in a volumetrically heated fluid with similar characteristics of an oxidic layer of a molten core in the lower head of the pressure vessel. The simulations were carried out in a square cavity with isothermal walls, for Rayleigh numbers (Ra) ranging from \$10$^9$\$ to \$10$^{11}$\$. Different turbulence models based on Reynolds Averaged Navier-Stokes equations were studied, such as the standard \$k-$\backslash$varepsilon\$, low-Reynolds-\$k-$\backslash$varepsilon\$ and Shear Stress Transport (SST), using the open-sourc-source Computational Fluid Dynamics (CFD) code - OpenFOAM (Open Field Operation and Manipulation). The results of the three turbulence models were compared versus the results of experimental correlations and other authors' simulations, and the conclusion was that the most promising model proves to be the SST, due to its accuracy and robustness.
Keywords:Natural Convection; Severe Accident; Turbulence Model.
Presentation Schedule: Wednesday, 16:00-16:20. Session: NCL-1. Presenter: Su Jian.
ENCIT2012-092 METHODOLOGY FOR RISK-BASED CONFIGURATION CONTROL OF NUCLEAR POWER PLANT OPERATION - PDF
Antonio Torres Valle, Instituto Superior de Tecnologías y Ciencias Aplicadas, Cuba
José de Jesús Rivero Oliva, UFRJ, Brazil
Abstract: The hazardous configurations control in Nuclear Power Plants is an application of a previous Probabilistic Safety Analysis (PSA). A more complete option would be the risk monitoring for the online detection of these configurations but expert personnel would be required to deal with the complexities of PSA and risk monitor. The paper presents a simpler but effective aproach: a method of configuration control, based on dependences matrixes. The algorithm is included in a computer code called SECURE A-Z. The configuration control is carried out in a qualitative way, without previous PSA results and not using a Risk Monitor. The simplicity of the method warrants its application to facilities where these tools have not been developed, allowing the detection of hazardous configurations during operation and increasing plant safety. This configuration control system was implemented in the Embalse Nuclear Power Plant in Argentina. The paper shows the application of the algorithm to the analysis of a simplified safety system.
Keywords:configuration control; Probabilistic Safety Analysis (PSA); dependences matrix; PSA applications; risk monitor.
Presentation Schedule: Wednesday, 16:20-16:40. Session: NCL-1. Presenter: Rivero Oliva, José de Jesús.
ENCIT2012-108 CHARACTERISATION OF RISING BUBBLES VELOCITY AND SHAPE IN A STAGNANT LIQUID VERTICAL COLUMN BY ULTRASONIC AND VISUALIZATION TECHNIQUES - PDF
Marcos Bertrand de Azevedo, CNEN, Brazil
Pedro Andrade Maia Vinhas, UFRJ, Brazil
José Luiz Horacio Faccini, CNEN, Brazil
Su Jian, UFRJ, Brazil
Abstract: The present paper reports a preliminary study of velocities of rising bubbles in a stagnant water vertical column. Rising bubble velocities were measure by using the pulse-echo ultrasonic technique and high speed digital camera visualization technique, which have been used in previous studies of two-phase gas-liquid flow in horizontal and slightly inclined circular tubes. Four types of rising bubbles were identified by flow visualization. We found that velocities of three types of rising bubbles agree reasonably well with available correlations, while velocities of one type of rising bubbles differ significantly from correlations.
Keywords:Ultrasonic Technique; Visualization Technique; Vertical Column.
Presentation Schedule: Thursday, 09:40-10:00. Session: NCL-2. Presenter: Marcos Bertrand de Azevedo.
ENCIT2012-096 TECHNICAL SPECIFICATIONS REVIEW OF NUCLEAR POWER PLANTS: A RISK-INFORMED EVALUATION - PDF
Pedro Luiz da Cruz Saldanha, Anna Letícia Sousa, CNEN, Brazil
Paulo Fernando Frutuoso e Melo, Juliana P. Duarte, UFRJ, Brazil
Abstract: The use of risk information by a regulatory body as part of an integrated decision making process addresses the way in which risk information is being used as part of an integrated process in making decisions about safety issues at nuclear plants - commonly referred to as risk-informed decision making. The risk-informed approach aims to integrate in a systematic manner quantitative and qualitative, deterministic and probabilistic safety considerations to obtain a balanced decision. PSA is a methodology that can be applied to provide a structured analysis process to evaluate the frequency and consequences of accidents scenarios in nuclear power plants. The Technical Specification,TS, are specifications regarding the characteristics of nuclear power plant (variables, systems or components) of overriding importance to nuclear safety and radiation protection, which is an integral part of plant operation authorization. The limiting conditions of operation, LCO, are the minimum levels of performance or capacity or operating system components required for the safe operation of nuclear plant, as defined in the technical specifications. The Maintenance Rule (MR) is a requirement established by the U. S. Nuclear Regulatory Commission (NRC) to check the effectiveness of maintenance carried out in nuclear plants, and to plant the configuration control. The control of plant configuration is necessary to verify the impact of the maintenance of a safety device out of service on plant safety. The Electric Power research Institute (EPRI) has assessed a role of probabilistic safety analysis in the regulation of nuclear power plant TS by the Report 1011758 with the following objectives: [1] to provide utilities with an approach for developing and implementing nuclear power station Risk-Managed Technical Specifications programs; and to complement and supplement existing successful configuration risk management applications such as the MR. The objective of this paper focuses on the evaluation of EPRI- Report 1011758 methodology with in risk informed decision making of changes to allowed outage times as a result of planned maintenance observing the MR requirements. The case study is related to planning maintenance of whose completion time exceeds the established Allowable Outage Time of TS
Keywords:Risk-Informed Regulation; Technical Specification; Maintenance Rule; Risk- Managed Technical Specifications.
Presentation Schedule: Thursday, 09:20-09:40. Session: NCL-2. Presenter: Pedro Luiz da Cruz Saldanha.
ENCIT2012-167 EXPERIMENTAL STUDY OF TWO-PHASE NATURAL CIRCULATION CIRCUIT - PDF
Wanderley Freitas Lemos, Su Jian, UFRJ, Brazil
José Luiz Horacio Faccini, CNEN, Brazil
Abstract: This paper reports an experimental study on the behavior of fluid flow in natural circulation under single-and two-phase flow conditions. The natural circulation circuit was designed based on concepts of similarity and scale in proportion to the actual operating conditions of a nuclear reactor. This test equipment has similar performance to the passive system for removal of residual heat presents in Advanced Pressurized Water Reactors (APWR). The experiment was carried out by supplying water to primary and secondary circuits, as well as electrical power resistors installed inside the heater. Power controller has available to ajsut the values for supply of electrical power resistors, in order to simulate conditions of decay of power from the nuclear reactor in steady state. Data acquisition system allows the measurement and control of the temperature at different points by means of thermocouples installed at several points along the circuit. The behavior of the phenomenon of natural circulation was monitored by a software with graphical interface, showing the evolution of temperature measurement points and the results stored in digital format spreadsheets. Besides, the natural circulation flow rate was measured by a flowmeter installed on the hot leg. A flow visualization technique was used the for identifying vertical flow regimes of two-phase natural circulation. Finally, the Reynolds Number was calculated for the establishment of a friction factor correlation dependent on the scale geometrical length, height and diameter of the pipe.
Keywords:natural circulation; residual heat removal; scaled installation; single-phase flow; two-phase flow.
Presentation Schedule: Thursday, 10:20-10:40. Session: NCL-2. Presenter: Wanderley Freitas Lemos.
ENCIT2012-261 CHEMICAL ANALYSIS OF THE ELEMENTS IN UZRNB ALLOY AT CDTN: PRELIMINARY INVESTIGATION - PDF
Wilmar Barbosa Ferraz, Helena Eugênia L. Palmieri, Sérgio Carneiro Reis, Ana Maria Matildes Santos, Adalberto Leles Souza, IME, Brazil
Abstract: The complete determination of major, minor, and impurity element contents in nuclear fuel is essential for quality assurance in the production of nuclear fuels. The control over all the stages of the development of nuclear fuel involves a combination of different analytical methods such as spectrometric methods. The goal of our investigation is to develop and evaluate procedures for the determination of main elements and carbon impurity present in some uranium alloys. In this paper the element contents in U2.5Zr7.5Nb and U3Zr9Nb alloys, in weight percent, were investigated by means of scanning electron microscopy with energy dispersive X-ray spectroscopy (SEM/EDS), inductively coupled plasma mass spectrometry (ICP-MS), wavelength dispersive fluorescence spectrometry (XRF/WDS) and energy-dispersive X-ray spectroscopy (EDX). The total carbon was determined using a carbon analyzer in which the sample is oxidized to carbon dioxide (IR absorption). It was observed a satisfactory correlation between the results obtained by employed methods.
Keywords:U-Zr-Nb; analytical methods; nuclear fuel.
Presentation Schedule: NO PRESENTATION.
ENCIT2012-263 EXPERIMENTAL AND NUMERICAL INVESTIGATION OF STRATIFIED GAS-LIQUID FLOW IN INCLINED CIRCULAR PIPES - PDF
José Luiz Horacio Faccini, Paulo Augusto Berquo de Sampaio, CNEN, Brazil
M. H. D. S. Botelho, J. S. Cunha Filho, Su Jian, UFRJ, Brazil
Abstract: In this paper, a stratified gas-liquid flow is experimentally and numerically investigated. Two measurement techniques, namely an ultrasonic technique and a visualization technique, are applied on an inclined circular test section using a fast single transducer pulse-echo technique and a high-speed camera. A numerical model is employed to simulate the stratified gas-liquid flow, formed by a system of non-linear differential equations consisting of the Reynolds averaged Navier-Stokes equations with the \&\#954;-\&\#969; turbulence model. The test section used in this work is comprised mainly of a transparent circular pipe with inner diameter 1 inch, and inclination angles varying from -2.5 to -10.0 degrees. Numerical solutions are obtained for the liquid height as a function of inclination angles, and compared with our own experimental data.
Keywords:stratified flow; ultrasonic technique; numerical model.
Presentation Schedule: Wednesday, 15:20-15:40. Session: NCL-1. Presenter: José Luiz Horacio Faccini.
ENCIT2012-018 DEVELOPMENT OF METHODS FOR MONITORING AND CONTROLLING POWER IN NUCLEAR REACTORS - PDF
Amir Zacarias Mesquita, Hugo Cesar Rezende, André Augusto Campagnole dos Santos, Vitor Vasconcelos Araújo Silva, IME, Brazil
Abstract: Redundancy and diversity are two important criteria for power measurement in nuclear reactors. Other criteria such as accuracy, reliability and response speed are also of major concern. Power monitoring of nuclear reactors is normally done by means of neutronic instruments, i.e. by the measurement of neutron flux. The greater the number of channels for power measuring the greater is the reliability and safety of reactor operations. The aim of this research is to develop new methodologies for on-line monitoring of nuclear reactor power using other reliable processes. One method uses the temperature difference between an instrumented fuel element and the pool water below the reactor core. Another method consists of the steady-state energy balance of the primary and secondary reactor cooling loops. A further method is the calorimetric procedure whereby a constant reactor power is monitored as a function of the temperature-rise rate and the system heat capacity. Another methodology, which does not employ thermal methods, is based on measurement of Cherenkov radiation produced within and around the core. The first three procedures, fuel temperature, energy balance and calorimetric, were implemented in the IPR-R1 TRIGA nuclear research reactor at Belo Horizonte (Brazil) and are the focus of the work described here. Knowledge of the reactor thermal power is very important for precise neutron flux and fuel element burnup calculations. The burnup is linearly dependent on the reactor thermal power and its accuracy is important in the determination of the mass of burned 235U, fission products, fuel element activity, decay heat power generation and radiotoxicity. The thermal balance method developed in this project is now the standard methodology used for IPR-R1 TRIGA reactor power calibration and the fuel temperature measuring is the most reliable way of on-line monitoring of the reactor power. This research project primarily aims at increasing the reliability and safety of nuclear reactors using alternative methods for power monitoring.
Keywords:Energy; Power; Nuclear reactor; Safety; Instrumentation.
Presentation Schedule: Wednesday, 14:40-15:00. Session: NCL-1. Presenter: Amir Zacarias Mesquita.
ENCIT2012-266 ANALYSIS OF POSTULATED LOSS OF COOLANT ACCIDENTS ON BRAZILIAN MULTIPURPOSE REACTOR USING RELAP5 - PDF
Humberto Vitor Soares, Antonella Lombardi Costa, Claubia Pereira Bezerra Lima, Maria Auxiliadora Fortini Veloso, UFMG, Brazil
Ivan Dionysio Aronne, IME, Brazil
Patrícia Amélia de Lima Reis, UFMG, Brazil
Abstract: The Brazilian Multipurpose Reactor (RMB) is currently being projected and several analyses are being carried out. It will be a 30 MW open pool multipurpose research reactor with a compact core using Materials Testing Reactor (MTR) type fuel assembly with planar plates. RMB will be cooled by light water and moderated by beryllium and heavy water. This work presents the calculations of steady state operation of RMB using the RELAP5 model and also three cases of loss of coolant accident (LOCA), in the reactor and service polls cooling system (RSPCS) inlet and two cases in the primary coolant system (PCS), inlet and outlet. In both cases the coolant pool level decreased until 7 m, keeping the core covered by water, but in different times. Natural circulation mode was established in the reactor pool and consequently the decay heat was removed keeping the integrity of the fuel elements.
Keywords:Research Reator; LOCA; RELAP5.
Presentation Schedule: Wednesday, 15:00-15:20. Session: NCL-1. Presenter: Humberto Vítor Soares.
ENCIT2012-270 ANALYSIS OF AN EXTREME LOSS OF COOLANT IN THE IPR-R1 TRIGA REACTOR USING A RELAP5 MODEL - PDF
Patrícia Amélia de Lima Reis, Antonella Lombardi Costa, Claubia Pereira Bezerra Lima, Maria Auxiliadora Fortini Veloso, UFMG, Brazil
Amir Mesquita Zacarias, IME, Brazil
Humberto Vitor Soares, UFMG, Brazil
Abstract: The RELAP5/MOD3.3 code has been applied for thermal hydraulic analysis of power reactors as well as nuclear research reactors with good predictions. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA have been validated for steady state and transient situations. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. In this work, an extreme transient case of loss of coolant accident (LOCA) has been simulated. For this type of analysis, the automatic scram of the reactor was not considered because the main aim was to verify the evolution of the fuel elements heating in the absence of coolant. The temperature evolutions are presented as well as an analysis about the temperature safety limits.
Keywords:LOCA; RELAP5; TRIGA IPR-R1.
Presentation Schedule: Thursday, 09:00-09:20. Session: NCL-2. Presenter: Patrícia Amélia de Lima Reis.
ENCIT2012-317 OVERVIEW OF LWR SEVERE ACCIDENT RESEARCH ACTIVITIES AT THE KARLSRUHE INSTITUTE OF TECHNOLOGY - PDF
Alexei Miassoedov, Giancarlo Albrecht, Jerzy Foit, Thomas Jordan, Martin Steinbrück, Juri Stuckert, Walter Tromm, Karlsruhe Institute of Technology, Germany
Abstract: The research activities in the light water reactor (LWR) severe accidents domain at Karlsruhe Institute of Technology (KIT) are concentrated on the in- and ex-vessel core melt behavior. The overall objective is to investigate the core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity and to the containment, and finally corium concrete interaction and corium coolability in the reactor cavity. The results of the experiments contribute to a better understanding of the core melt sequences and thus improve safety of existing and, in the long-term, of future reactors by severe accident mitigation measures and by safety installations where required. This overview paer describes the experimental facilities used at KIT for severe accident research and gives an overview of the main directions and objectives of the R\&D work.
Keywords:LWR; reactor safety; severe accidents; corium behavior.
Presentation Schedule: Wednesday, 14:20-14:40. Session: NCL-1. Presenter: Alexei Miassoedov.
ENCIT2012-345 ANGRA 2 SMALL BREAK LOCA FLOW REGIME IDENTIFICATION THROUGH RELAP5 CODE - PDF
Marcelo da Silva Rocha, Gaianê Sabundjian, Antônio Belchior-Jr, Delvonei Alves de Andrade, Walmir Maximo Torres, Thadeu das Neves Conti, Luiz Alberto Macedo, Pedro Ernesto Umbehaun, Roberto Navarro de Mesquita, Paulo Henrique F. Masotti, CNEN, Brazil
Abstract: The purpose of this paper is to identify the flow regimes in the core of Angra 2 nuclear reactor with RELAP5/MOD3.2.gama code (RELAP5, 2001). The postulated accident is the loss of coolant through a small break in the primary circuit (SBLOCA), which is described in Chapter 15 of the Final Safety Analysis Report of Angra 2 - FSAR (ETN, 2006). As the primary circuit pressure decreases due to the loss of coolant, several alternating two phase flow regimes are established in the primary circuit. This paper analyses the coolant two-phase flow behavior in the nuclear reactor core during the postulated accident.
Keywords:RELAP5; Small Break LOCA; Flow Regimes.
Presentation Schedule: Wednesday, 15:40-16:00. Session: NCL-1. Presenter: Marcelo da Silva Rocha.