Assessment of the Integrity of the Channel Shelter of the Experimental Nuclear Reactor LVR 15

2009 ◽  
Author(s):  
Miroslav Svrcek ◽  
Ivan Krasny ◽  
Jan Lestina
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Nuclear Research ◽  
Point Of View ◽  
Internal Diameter ◽  
Neutron Absorption ◽  
Nuclear Research Institute ◽  
Channel Water ◽  
Channel Tube ◽  
The Stability

The experimental reactor LVR 15 is operated in Nuclear Research Institute Rez. A pressurized experimental channel forming a field tube is installed in the reactor core. The operating parameters of the channel water are 16MPa and 290°C. From the point of view of operation (neutron absorption) the channel is located in a quadratic shelter 68×68mm with internal diameter 65mm (see Fig. 2). Based on the recommendation of IAEA INSAR mission, the integrity of the shelter must be conserved if a through wall crack in the outer channel tube is postulated, moreover the stability of the through wall crack must be guaranteed. Solutions of both situations are presented in the paper.

scholarly journals Thermal-hydraulic behavior of physical quantities at critical velocities in a nuclear research reactor core channel using plate type fuel

2012 ◽  
Vol 27 (3) ◽  
pp. 229-238
Author(s):  
Ali Sidi ◽  
Zaki Boudali ◽  
Rachid Salhi
Keyword(s):  
Heat Flux ◽  
Critical Velocity ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Thermal Characteristics ◽  
Nuclear Research ◽  
Coolant Fluid ◽  
Critical Velocities ◽  
Set Up ◽  
Physical Quantities

The thermal-hydraulic study presented here relates to a channel of a nuclear reactor core. This channel is defined as being the space between two fuel plates where a coolant fluid flows. The flow velocity of this coolant should not generate vibrations in fuel plates. The aim of this study is to know the distribution of the temperature in the fuel plates, in the cladding and in the coolant fluid at the critical velocities of Miller, of Wambsganss, and of Cekirge and Ural. The velocity expressions given by these authors are function of the geometry of the fuel plate, the mechanical characteristics of the fuel plate?s material and the thermal characteristics of the coolant fluid. The thermal-hydraulic study is made under steady-state; the equation set-up of the thermal problem is made according to El Wakil and to Delhaye. Once the equation set-up is validated, the three critical velocities are calculated and then used in the calculations of the different temperature profiles. The average heat flux and the critical heat flux are evaluated for each critical velocity and their ratio reported. The recommended critical velocity to be used in nuclear channel calculations is that of Wambsganss. The mathematical model used is more precise and all the physical quantities, when using this critical velocity, stay in safe margins.

scholarly journals Physical and Numerical Difficulties in Computer Modelling of Pellet-Cladding Contact Problems for Burned-Up Fuel

10.14311/756 ◽  
2005 ◽  
Vol 45 (5) ◽  
Author(s):  
M. Dostál ◽  
J. Zymák ◽  
M. Valach
Keyword(s):  
Electricity Market ◽  
Computer Modelling ◽  
Nuclear Research ◽  
Contact Problems ◽  
Point Of View ◽  
3D Fem ◽  
Nuclear Research Institute ◽  
Electricity Grid ◽  
Deregulated Electricity Market ◽  
Safety Margins

The importance of fuel reliability is growing due to the deregulated electricity market and the demands on operability and availability to the electricity grid of nuclear units. Under these conditions of fuel exploitation, the problems of PCMI (Pellet-Cladding Mechanical Interaction) are very important from the point of view of fuel rod integrity and reliability. Severe loading is thermophysically and mechanically expressed as a greater probability of cladding failure especially during power maneuvering. We have to be able to make a realistic prediction of safety margins, which is very difficult by using computer simulation methods. NRI (Nuclear Research Institute) has recently been engaged in developing 2D and 3D FEM (Finite Element Method) based models dealing with this problem. The latest effort in this field has been to validate 2D r-z models developed in the COSMOS/M system against calculations using the FEMAXI-V code. This paper presents a preliminary comparison between classical FEM based integral code calculations and new models that are still under development. The problem has not been definitely solved. The presented data is of a preliminary nature, and several difficult problems remain to be solved. 

Implications of Quench Bundle Test Simulations for Plant Applications With MELCOR Code

2008 ◽  
Author(s):  
Jirˇi´ Duspiva
Keyword(s):  
Hydrogen Generation ◽  
Reactor Core ◽  
Nuclear Research ◽  
Failure Criteria ◽  
Heat Losses ◽  
Onset Temperature ◽  
Severe Accident ◽  
Experimental Program ◽  
Correct Treatment ◽  
Nuclear Research Institute

Reflooding of an uncovered and overheated core is one of the most expected measures during the early phase of a severe accident progression. The capability of analytical tools to model correctly processes during independent sub-phases of this accident progression is on different levels, and the most concentrated effort is focused on the study of fuel cladding quenching topic. The main objective of the Forschungszentrum Karlsruhe (Germany) Quench experimental program is to investigate the hydrogen generation during the reflooding of overheated core. The Nuclear Research Institute Rez (hereafter NRI) contributes to this program by the analytical simulation of the Quench bundle experiments with the MELCOR code mainly and ICARE2 code as well since 2000. The NRI analyses covered the Quench-01, Quench-03 and Quench-06 with version MELCOR 1.8.5 (including reflood model), and Quench-01 and Quench-11 tests with the latest version MELCOR 1.8.6. The tests Quench-01 and Quench-06 were characteristic of the lower reflooding onset temperature (1900–2050 K), and the tests Quench-03 and Quench-11 had high reflooding onset temperature (> 2350 K) with the fast heat-up phase before reflooding. The integral code MELCOR is capable to sufficiently predict the heat up and reflooding phases for the tests with the lower onset temperature, but the Quench-03 test with high onset temperature was not predicted correctly starting in the heat-up phase and also the hydrogen generation during bundle reflooding was significantly underpredicted. The interpretation of the Quench-03 test with ICARE2 code confirmed the MELCOR results, but its more detail description enabled to identify a cause. The underestimation of the bundle temperatures is influenced by the shroud insulation behaviour, which could be interpreted as the reduction of heat losses through shroud and simultaneously more intensive heat-up of a bundle. The phenomenon, which causes this behaviour, is not yet known, and is not important for plant applications, although it is relevant for the test interpretation. Also the interpretation of the Quench-11 showed an importance of correct modeling of the bundle shroud, which is in the MELCOR 1.8.6 simulated by newly implemented independent component. The Quench-11 test analysis showed simplified failure criteria for the cases when the shroud is not supported by the former components, which are assumed in the basic MELCOR model fundamentals. Concerning the plant application of the MELCOR code for the scenario with the overheated core reflooding, the user has to give attention on the correct treatment of heat losses through shroud and its degradation modeling, mainly if the shroud supporting formers absent in the core periphery design. Hydrogen generation during reflooding is predicted well for lower onset temperatures (below 2100 K) and slightly underpredicted for higher onset temperatures. It is important to take these code capabilities into account because a reflooding of the reactor core will consist of varied reflooding onset temperatures.

Reevaluation of Spectral Parameters and Neutron Fluxes in IC-7 Irradiation Channel of TRIGA MARK I IPR-R1 Research Nuclear Reactor

2021 ◽  
Author(s):  
Radojko Jacimovic ◽  
Maria Angela de Barros Correia Menezes
Keyword(s):  
Nuclear Reactor ◽  
Epithermal Neutron ◽  
Research Reactor ◽  
Nuclear Research ◽  
Irradiation Channel ◽  
Spectral Parameters ◽  
K0 Method ◽  
Nuclear Research Reactor ◽  
Research Nuclear Reactor ◽  
The Stability

Abstract The core configuration of the TRIGA MARK I IPR-R1 nuclear research reactor, Brazil, has been modified six times since the first criticality and the neutron fluxes have been determined using experimental and semi theoretical methodologies determining the neutron fluxes in different irradiation channels and devices, applying different procedures and materials. This reactor operates at 100 kW, however, after new configuration for 250 kW in 2001, the carousel no longer rotates during irradiations aiming at preserving the rotation mechanism. In 2003, the spectral parameters were determined experimentally by the "Cd-ratio for multi-monitor" in five specific channels aiming at the application of NAA k0-standardized method. The determinations were repeated applying the same procedure in 2016, 2018 and 2019. Values for thermal and epithermal neutron fluxes as well as f and a spectral parameters were determined. The experimental results for CRM BCR-320R were calculated by the k0-method of NAA, using the spectral parameters for irradiation channel IC-7 obtained in 2003, 2016, 2018 and 2019 and evaluated by En-score. The values showed that the differences in the results compared to those in 2003 were lower than 2.5%, inside the uncertainty of the method. It shows that the k0-method installed in CDTN is reliable and useful for various purposes. The results of the spectral parameter f presented small differences, in a period of 16 years, pointing out the stability of operation of the reactor TRIGA MARK I IPR-R1.

scholarly journals Physical Ageing of The Research Reactor Core Structural Materials Due To Neutron Irradiation Exposure: A Review

2017 ◽  
Vol 18 (2) ◽  
pp. 93
Author(s):  
Julwan Hendry Purba
Keyword(s):  
Neutron Irradiation ◽  
Nuclear Reactor ◽  
Research Reactor ◽  
Reactor Core ◽  
Structural Materials ◽  
Industrial Applications ◽  
Physical Ageing ◽  
Functional Capabilities ◽  
Wide Range ◽  
The Stability

A research reactor (RR) is a nuclear reactor that has function to generate and utilize neutron flux and radiation ionization for research purposes and industrial applications. More than 60% of current operating RRs have been operated for 30 years or more. As the time passes, the functional capabilities of structures, systems and components (SSCs) of those RRs deteriorate by physical ageing, which can be caused by neutron irradiation exposure such as irradiation induced dislocation and microstructural changes. To extend the lifetime and/or to avoid unplanned outages, ageing on the safety related SSCs of RRs need to be properly managed. An ageing management is a strategy to engineer, operate, maintenance, and control SSC degradation within acceptablelimits. The purpose of this study is to review physical ageing of the core structural materials of the RRs caused by neutron irradiation exposure. In order to achieve this objective, a wide range of literatures are reviewed. Comprehensive discussions on irradiation behaviors are limited only on reactor vessel and core support structure materials made from zirconium and beryllium as well as their alloys, which are widely used in RRs. It is found that the stability of the mechanical properties of zirconium and beryllium as well as their alloys was mostly affected by the neutron fluences and temperatures.

A Computer Program Predicting Steady-State Performance of a Nuclear Research Reactor’s Cooling System

2002 ◽  
Author(s):  
Kamel Sidi Ali
Keyword(s):  
Computer Program ◽  
Nuclear Reactor ◽  
Cooling System ◽  
Reactor Core ◽  
Nuclear Research ◽  
Optimal Number ◽  
Climate Conditions ◽  
Current Climate ◽  
Connected Set ◽  
Nuclear Reactor Core

The performances of a nuclear reactor are directly affected by its cooling system, especially when it uses wet towers to evacuate the heat generated in the nuclear reactor core. Failure of the cooling system can yield very serious damages to most of the components of the nuclear reactor core. In this work, a computer program simulating the thermal behavior of a nuclear research reactor’s cooling system is presented. Starting from the proposed start-up data of the reactor, the program predicts the cooling capacity of the nuclear reactor while taking into account the current climate conditions and also monitors the behavior of the thermal equipments involved in this process and this for different levels of power. The proposed simulation is based on a set of heat transfer equations representing all the equipments making up the cooling system up to the nuclear reactor core. Owing to the proposed inter-connected set of equations used to predict the thermal behaviour of the system, this program allows the user to modify at will a specified parameter and study the induced resulting effects on the rest of the system. The computer program developed has been experimentally validated on an operational system generating 6.8 MW and the obtained results are in good agreement with experiment. The results produced by the program concern the capacity of the cooling system to evacuate all the heat generated in the nuclear reactor core while taking into account the current climate conditions, the determination of the optimal number of thermal equipments that need to be engaged, the monitoring of the reactor core’s entry end exit temperatures as well as the temperatures of all the components of the cooling system. Moreover, the program gives all the characteristics of air at the exit of the cooling towers and the loss of water due to the cooling process.

In situ TEM studies of irradiation effects for materials improvement

1991 ◽  
Vol 49 ◽  
pp. 452-453
Author(s):  
Charles W. Allen
Keyword(s):  
Nuclear Reactor ◽  
Austenitic Stainless Steels ◽  
High Energy ◽  
Point Of View ◽  
In Situ Tem ◽  
Irradiation Effects ◽  
Tem Analysis ◽  
Radiation Induced ◽  
Analytical Tools

Irradiation effects studies employing TEMs as analytical tools have been conducted for almost as many years as materials people have done TEM, motivated largely by materials needs for nuclear reactor development. Such studies have focussed on the behavior both of nuclear fuels and of materials for other reactor components which are subjected to radiation-induced degradation. Especially in the 1950s and 60s, post-irradiation TEM analysis may have been coupled to in situ (in reactor or in pile) experiments (e.g., irradiation-induced creep experiments of austenitic stainless steels). Although necessary from a technological point of view, such experiments are difficult to instrument (measure strain dynamically, e.g.) and control (temperature, e.g.) and require months or even years to perform in a nuclear reactor or in a spallation neutron source. Consequently, methods were sought for simulation of neutroninduced radiation damage of materials, the simulations employing other forms of radiation; in the case of metals and alloys, high energy electrons and high energy ions.

scholarly journals Suboptimal Disturbance Observer Design Using All Stabilizing Q Filter for Precise Tracking Control

Mathematics ◽  
2020 ◽  
Vol 8 (9) ◽  
pp. 1434 ◽  
Author(s):  
Wonhee Kim ◽  
Sangmin Suh
Keyword(s):  
Design Method ◽  
Industrial Applications ◽  
Point Of View ◽  
Observer Design ◽  
Linear Matrix ◽  
External Disturbances ◽  
Closed Loop Systems ◽  
Control Target ◽  
The Stability ◽  
Unified Index

For several decades, disturbance observers (DOs) have been widely utilized to enhance tracking performance by reducing external disturbances in different industrial applications. However, although a DO is a verified control structure, a conventional DO does not guarantee stability. This paper proposes a stability-guaranteed design method, while maintaining the DO structure. The proposed design method uses a linear matrix inequality (LMI)-based H∞ control because the LMI-based control guarantees the stability of closed loop systems. However, applying the DO design to the LMI framework is not trivial because there are two control targets, whereas the standard LMI stabilizes a single control target. In this study, the problem is first resolved by building a single fictitious model because the two models are serial and can be considered as a single model from the Q-filter point of view. Using the proposed design framework, all-stabilizing Q filters are calculated. In addition, for the stability and robustness of the DO, two metrics are proposed to quantify the stability and robustness and combined into a single unified index to satisfy both metrics. Based on an application example, it is verified that the proposed method is effective, with a performance improvement of 10.8%.

scholarly journals No-z Model for Magnetic Fields of Different Astrophysical Objects and Stability of the Solutions

Data ◽  
2021 ◽  
Vol 6 (1) ◽  
pp. 4
Author(s):  
Evgeny Mikhailov ◽  
Daniela Boneva ◽  
Maria Pashentseva
Keyword(s):  
Magnetic Field ◽  
Magnetic Fields ◽  
Large Scale ◽  
Point Of View ◽  
Accretion Discs ◽  
Wide Range ◽  
Astrophysical Objects ◽  
Alpha Effect ◽  
The Stability ◽  
The Magnetic Field

A wide range of astrophysical objects, such as the Sun, galaxies, stars, planets, accretion discs etc., have large-scale magnetic fields. Their generation is often based on the dynamo mechanism, which is connected with joint action of the alpha-effect and differential rotation. They compete with the turbulent diffusion. If the dynamo is intensive enough, the magnetic field grows, else it decays. The magnetic field evolution is described by Steenbeck—Krause—Raedler equations, which are quite difficult to be solved. So, for different objects, specific two-dimensional models are used. As for thin discs (this shape corresponds to galaxies and accretion discs), usually, no-z approximation is used. Some of the partial derivatives are changed by the algebraic expressions, and the solenoidality condition is taken into account as well. The field generation is restricted by the equipartition value and saturates if the field becomes comparable with it. From the point of view of mathematical physics, they can be characterized as stable points of the equations. The field can come to these values monotonously or have oscillations. It depends on the type of the stability of these points, whether it is a node or focus. Here, we study the stability of such points and give examples for astrophysical applications.

scholarly journals Photoluminescent Coordination Polymers Based on Group 12 Metals and 1H-Indazole-6-Carboxylic Acid

Inorganics ◽  
2021 ◽  
Vol 9 (3) ◽  
pp. 20
Author(s):  
Antonio A. García-Valdivia ◽  
Estitxu Echenique-Errandonea ◽  
Gloria B. Ramírez-Rodríguez ◽  
José M. Delgado-López ◽  
Belén Fernández ◽  
...  
Keyword(s):  
Carboxylic Acid ◽  
Coordination Polymers ◽  
Density Functional ◽  
Free Ligand ◽  
Emission Spectra ◽  
Point Of View ◽  
Theoretical Point ◽  
Fourier Transformed Infrared Spectroscopy ◽  
3D Network ◽  
The Stability

Two new coordination polymers (CPs) based on Zn(II) and Cd(II) and 1H-indazole-6-carboxylic acid (H2L) of general formulae [Zn(L)(H2O)]n (1) and [Cd2(HL)4]n (2) have been synthesized and fully characterized by elemental analyses, Fourier transformed infrared spectroscopy and single crystal X-ray diffraction. The results indicate that compound 1 possesses double chains in its structure whereas 2 exhibits a 3D network. The intermolecular interactions, including hydrogen bonds, C–H···π and π···π stacking interactions, stabilize both crystal structures. Photoluminescence (PL) properties have shown that compounds 1 and 2 present similar emission spectra compared to the free-ligand. The emission spectra are also studied from the theoretical point of view by means of time-dependent density-functional theory (TD-DFT) calculations to confirm that ligand-centred π-π* electronic transitions govern emission of compound 1 and 2. Finally, the PL properties are also studied in aqueous solution to explore the stability and emission capacity of the compounds.

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