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[en] This work presents the experiments and theoretical analysis to determine the temperature parameter of the uranium zirconium hydride fuel elements, used in the TRIGA IPR-R1 Research Nuclear Reactor. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were evaluated experimentally. It was also presented a correlation for the gap conductance between the fuel and the cladding. In the case of nuclear fuels the heat parameters become functions of the irradiation as a result of change in the chemical and physical composition. The value of the heat transfer coefficients should be determined experimentally. (author)
[en] One of the most important subjects in nuclear reactor safety analysis is the reactor core rewetting after a Loss-of-Coolant Accident (LOCA) in a Light Water Reactor LWR. Several codes for the prediction of the rewetting evolution are under development based on experimental results. In a Pressurized Water Reactor (PWR) the reflooding phase of a LOCA is when the fuel rods are rewetted from the bottom of the core to its top after having been totally uncovered and dried out. Out-of-pile reflooding experiments performed with electrical heated fuel rod simulators show different quench behavior depending the rods geometry. A test facility for rewetting experiments (ITR - Instalacao de Testes de Remolhamento) has been constructed at the Thermal Hydraulics Laboratory of the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), with the objective of performing investigations on basic phenomena that occur during the reflood phase of a LOCA in a PWR, using tubular and annular test sections. This paper presents the design aspects of the facility, and the current stage of the works. The mechanical aspects of the installation as its instrumentation are described. Two typical tests are presented and results compered with theoretical calculations using computer code. (author)
[en] This work presents the experiments analysis to determine the temperature distribution in the -IPR-R1 TRIGA Research Nuclear Reactor of the CDTN, Belo Horizonte, Minas Gerais. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. A data acquisition and processing system and a software were developed to help the investigation. (author)
[en] A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)
[en] Experimental studies had been performed in the TRIGA Research Nuclear Reactor of CDTN/CNEN to find out the its thermal hydraulic parameters. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power in order to assess the thermal hydraulic performance of the core. The heat generated by nuclear fission in the reactor core is transferred from fuel elements to the cooling system through the fuel-cladding (gap) and the cladding to coolant interfaces. As the reactor core power increases the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. This paper presents the uncertainty analysis in the results of the thermal hydraulics experiments performed. The methodology used to evaluate the propagation of uncertainty in the results was done based on the pioneering article of Kline and McClintock, with the propagation of uncertainties based on the specification of uncertainties in various primary measurements. The uncertainty analysis on thermal hydraulics parameters of the CDTN TRIGA fuel element is determined, basically, by the uncertainty of the reactor's thermal power. (author)
[en] The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)
[en] In order to study the safety aspects connected with the permanent increase of the maximum steady state power of the IPR-R1 TRIGA Reactor of the Nuclear Technology Development Center (CDTN), experimental measurements were done with the reactor operating at power levels of 265 kW and 105 kW, with the pool forced cooling system turned off. A number of parameters were measured in real-time such as fuel and water temperatures, radiation levels, reactivity, and influence of cooling system. Information on all aspects of reactor operation was displayed on the Data Acquisition System (DAS) shown the IPR-R1 online performance. The DAS was developed to monitor and record all operational parameters. Information displayed on the monitor was recorded on hard disk in a historical database. This paper summarizes the behavior of some operational parameters, and in particular, the evolution of the temperature in the fuel element centerline positioned in the core hottest location. The natural circulation test was performed to confirm the cooling capability of the natural convection in the IPR-R1 reactor. It was confirmed that the IPR-R1 has capability of long-term core cooling by natural circulation operating at 265 kW. The measured maximum fuel temperature of about 300 deg C was lower than the operating limit of 550 deg C. Fuel, channel and pool temperatures in natural circulation depend on reactor power, as well as on environment temperature. (author)
[en] In the thermal hydraulic experiments to determinate parameters of heat transfer, where fuel rod simulators are heated by electric current, the preservation of the simulators are essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The departure from nucleate boiling (DNB) happens in the area of low steam quality when there is nucleus formation of bubbles. This result in a departure from nucleate boiling in which steam bubbles no longer break away from the solid surface of the channel, bubbles dominate the channel or surface, and the heat flux dramatically decreases. Vapor essentially insulates the bulk liquid from the hot surface. At this time, the small increase in the heat flux or in the inlet temperature of the cooler in the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detection of critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting of critical heat flux in nuclear simulators heated by electric current in open pool. (author)
[en] Highlights: • This work analyzed the models that governing the rewetting phenomenon in PWR reactor. • An analysis of temperature evolution and of reweting front was made for two experiments. • The results obtained with REWET code were compared with the experimental results. • The result simulated for the rewetting time came closer to the experimental results. - Abstract: The safety of nuclear power plants is determined by their protection against the possible outcomes of postulated accidents. One of the most important accidents is the loss of coolant in the core (Loss-of-Coolant Accident – LOCA). A process of fundamental importance in the event of a LOCA in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Centre (CDTN) has been developing programmes since the 1970s to make Brazil independent in the field of reactor safety analysis. To that end, a Rewetting Test Facility (ITR in Portuguese) was designed, assembled and commissioned. This facility aims at investigating the phenomena involved in the thermal hydraulic reflood phase of a LOCA in a PWR nuclear reactor. The objective of this work is the analysis of physical and mathematical models that govern the rewetting phenomenon. Thus, a simulation code was developed for the Rewetting Test Facility (ITR), which represents a PWR core cooling channels. The thermohydraulic code was named Rewet. The results obtained with Rewet code were compared with the experimental results of the ITR and with the results of the Hydroflut code, the program used until then. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the rewetting front for two typical tests using the two-code calculation and experimental results. In all cases, there was a better adjustment of the Rewet results in relation to those of the Hydroflut. The result simulated by the Rewet code for the rewetting time also came closer to the experimental results more than that obtained with the Hydroflut code.
[en] Highlights: ► Thermal stratification in PWR piping was numerically and experimentally studied. ► Verification and Validation was performed according to ASME V and V 20 standard. ► Numerical and experimental transient temperature behavior showed good agreement. ► V and V enable the objective identification of the strengths and weaknesses of the model. - Abstract: Thermal stratification and striping are observed in many piping systems including those of nuclear power plants. Periodic occurrences of these thermal transients lead to fatigue and may induce undesirable failures and deformations to the piping. The Thermal Hydraulic Laboratory of the Centro de Desenvolvimento da Tecnologia Nuclear/Comissão Nacional de Energia Nuclear (CDTN/CNEN) conducts an experimental and numerical project simulating the thermal stratified flows in piping systems of pressurized water reactors (PWR) to obtain some understanding on these phenomena. Experiments were carried out in a test section simulating the steam generator injection nozzle of a PWR. A numerical simulation of one experiment was performed with the commercial finite volume Computational Fluid Dynamic code CFX 13.0. A vertical symmetry plane along the pipe was adopted to reduce the geometry in one half, reducing mesh element size and minimizing processing time. The RANS two equations RNG k–ε turbulence model with scalable wall function and the full buoyancy model were used in the simulation. In order to properly evaluate the numerical model it was performed a Verification and Validation (V and V) process. Numerical uncertainties due to mesh refinement and time step were evaluated. This validation process showed the great importance of a proper quantitative evaluation of numerical results. In past studies qualitative evaluations were considered enough and numerical results like the one presented here could be considered satisfactory for the prediction of thermal stratified flows. However, with the present V and V study it was possible to identify objectively the strengths and weaknesses of the model.