scholarly journals Prediction of thermophysical characteristics of fuel rods based on calculations

2021 ◽  
Vol 7 (2) ◽  
pp. 79-86
Author(s):  
Stepan Lys ◽  
◽  
Igor Galyanchuk ◽  
Tetiana Kovalenko
Keyword(s):  
Reactor Core ◽  
Operating Conditions ◽  
Design Criteria ◽  
Gas Release ◽  
Fuel Cladding ◽  
Thermophysical Characteristics ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Main Factors

The paper analyzes operating conditions, thermophysical characteristics were calculated as applied to WWER-1000 fuel rods in a four-year cycle for unified core. The paper gives a summary of models for calculating gas release, pressure of gases within fuel rod cladding, fuel swelling and thermal conductivity, fuel-cladding gap conductance. The thermophysical condition of fuels in a reactor core is one of the main factors that determine their serviceability. The stress-strained condition of fuel claddings under design operating conditions is closely related to fuel rod temperature, swelling, gas release from fuel pellets and the mode in which they change during the cycle and transients. Aside from this, those parameters are an independent goal of studies since their ultimate values are governed by the system of design criteria.

scholarly journals CALCULATION OF STRESS AND DEFORMATION IN FUEL ROD CLADDING DURING PELLET-CLADDING INTERACTION

2015 ◽  
Vol 55 (6) ◽  
pp. 384 ◽  
Author(s):  
Dávid Halabuk ◽  
Jiří Martinec
Keyword(s):  
Reactor Core ◽  
Fuel Pellet ◽  
Deformation Analysis ◽  
Main Function ◽  
Fuel Cladding ◽  
Fuel Rods ◽  
Radial Temperature ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Stress And Deformation

The elementary parts of every fuel assembly, and thus of the reactor core, are fuel rods. The main function of cladding is hermetic separation of nuclear fuel from coolant. The fuel rod works in very specific and difficult conditions, so there are high requirements on its reliability and safety. During irradiation of fuel rods, a state may occur when fuel pellet and cladding interact. This state is followed by changes of stress and deformations in the fuel cladding. The article is focused on stress and deformation analysis of fuel cladding, where two fuels are compared: a fresh one and a spent one, which is in contact with cladding. The calculations are done for 4 different shapes of fuel pellets. It is possible to evaluate which shape of fuel pellet is the most appropriate in consideration of stress and deformation forming in fuel cladding, axial dilatation of fuel, and radial temperature distribution in the fuel rod, based on the obtained results.

Identification of Non-Linear Vibration Parameters of a Nuclear Fuel Rod

2018 ◽  
Author(s):  
Marco Amabili ◽  
Prabakaran Balasubramanian ◽  
Giovanni Ferrari ◽  
Stanislas Le Guisquet ◽  
Kostas Karazis ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Harmonic Balance Method ◽  
Resonance Phenomenon ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Safety Margins ◽  
Fuel Assemblies ◽  
Non Linear

In Pressurized Water Reactors (PWR), fuel assemblies are composed of fuel rods, long slender tubes filled with uranium pellets, bundled together using spacer grids. These structures are subjected to fluid-structure interactions, due to the flowing coolant surrounding the fuel assemblies inside the core, coupled with large-amplitude vibrations in case of external seismic excitation. Therefore, understanding the non-linear response of the structure and, particularly, its dissipation, is of paramount importance for the choice of safety margins. To model the nonlinear dynamic response of fuel rods, the identification of nonlinear stiffness and damping parameters is required. The case of a single fuel rod with clamped-clamped boundary conditions was investigated by applying harmonic excitation at various force levels. Different configurations were implemented testing the fuel rod in air and in still water; the effect of metal pellets simulating nuclear fuel pellets inside the rods was also recorded. Non-linear parameters were extracted from some of the experimental response curves by means of a numerical tool based on the harmonic balance method. The axisymmetric geometry of fuel rods resulted in the presence of a one-to-one internal resonance phenomenon, which has to be taken into account modifying accordingly the numerical identification tool. The internal motion of fuel pellets is a cause of friction and impacts, complicating further the linear and non-linear dynamic behavior of the system. An increase of the equivalent viscous-based modal damping with excitation amplitude is often shown during geometrically non-linear vibrations, thus confirming previous experimental findings in the literature.

Projection of the Neutronic and Thermal Fuel Rod Behavior in a BWR

2004 ◽  
Author(s):  
Hector Hernandez Lopez ◽  
Javier Ortiz Villafuerte
Keyword(s):  
Fuel Assembly ◽  
Mechanical Performance ◽  
Nuclear Research ◽  
Operating Conditions ◽  
Fast Neutrons ◽  
Two Phase ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Thermal Limits ◽  
Operation Conditions

Currently, at the Instituto Nacional de Investigaciones Nucleares (National Institute for Nuclear Research) in Mexico, it is being developed a computational code for evaluating the neutronic, thermal and mechanical performance of a fuel element at several different operation conditions. The code is referred as to MCTP (Multigrupos con Temperaturas y Potencia), and is benchmarked against data from the Laguna Verde Nuclear Power Plant (LVNPP). In the code, the neutron flux is approximated by six groups of energy: one group in the thermal region (E < 0.625 eV), four in the resonances region (0.625 eV < E < 0.861 MeV), and one group in the fast region (E > 0.861 MeV). Thus, the code is able to determine the damage to the cladding due to fast neutrons. The temperature distribution is approximated in both axial and radial directions taking into account the changes in the coolant density, for both the single and two-phase regions in a BWR channel. It also considerate the changes in the thermal conductivity of all materials involved for the temperature calculations, as well as the temperature and density effects in the neutron cross sections. In the code, fuel rod burnup is evaluated. Also, plutonium production and poison production from fission. In this work, the neutronic and thermal performance of fuel rods in a 10×10 fuel assembly is evaluated. The fuel elements have a content of 235U. The fuel assembly was introduced to the unit 1 of LVNPP reactor core in the cycle 9 of operation, and will stay in during three cycles. In the analysis of fuel rod performance, the operating conditions are those for the cycle 9 and 10, whereas for the current cycle (cycle 11) the reactor is projected to operate during 460 days. The analysis for cycle 11 uses the actual location of the fuel assembly that will have in the core. The results show that the fuel rods analyzed did not reach the thermal limits during the cycles 9 and 10, as expected, and for cycle 11 the same thermal limits are not predicted to be reached.

The Research on Core Melting Process: Oxidation

2013 ◽  
Author(s):  
Jun Wang ◽  
Wenxi Tian ◽  
Jianan Lu ◽  
Yingying Ma ◽  
Guanghui Su ◽  
...  
Keyword(s):  
Natural Circulation ◽  
Reactor Core ◽  
Melting Process ◽  
Coolant Pump ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Design Basis ◽  
Loss Of Coolant ◽  
Design Basis Accident ◽  
Station Blackout

Beyond-design basis accidents in the AP1000 may result in reactor core melting and are therefore termed core melt accidents. The aim of this work is to develop a code to calculate and analyze the oxidation of a single fuel rod with total failures of engineered safeguard systems under a certain beyond-design basis accident such as a gigantic earthquake which can result in station blackout and then total loss of coolant flow. Using the code, the responses of the most dangerous fuel rod in the AP1000 were calculated under the accident. A discussion involving fuel pellets melting, cladding rupture and oxidation, and hydrogen production then was carried out, focused on DNBR during coolant pump coastdown, the cladding intactness under different flow rates in natural circulation, and the delay effect on cladding rupture due to cladding oxidation. By the analysis of calculated results, several suggestions on guaranteeing the security of fuel rods were provided.

scholarly journals MODEL FOR CALCULATING THE TEMPERATURE IN THE FUEL RODS OF THE FA-X FUEL ASSEMBLY PRODUCED FOR SUBCRITICAL INSTILLATION AND REACTOR WWR-M

2020 ◽  
Vol 2 (61) ◽  
pp. 31-41
Author(s):  
I. Chernov ◽  
◽  
А. Кushtym ◽  
Keyword(s):  
Temperature Distribution ◽  
Fuel Element ◽  
Research Reactor ◽  
Operating Conditions ◽  
Maximum Temperature ◽  
Software Module ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Cross Section Area ◽  
Ceramic Fuel

The TVS-X fuel rod model designed by NSC KIPT as an alternative fuel for subcritical assembly (SCA, KIPT, Kharkov) and research reactor (WWR-M, INR, Kiev) is described. The model is a program that allows calculating the temperature distribution on the radius and height of the fuel element containing both uranium oxide pellets and dispersion fuel based on the UO2+Al composition with different contents of the fuel phase, as well as the different geometric characteristics of the fuel element and the values of the coolant parameters: the temperature at the entrance to the hydraulic channel and the coolant speed. Comparative calculations of temperature distribution during operation are carried out. As a result, it has been shown that for conditions of operation in the SCA (linear power of fuel rod is 2.62 kW/m), the fuel center temperature reaches ~140 °C for UO2 and ~112 °C for the UO2+Al composition. For operating conditions in the WWR-M reactor (linear power of fuel rod is 12.1 kW/m), the fuel center temperature reaches ~626 °C for ceramic (UO2) and ~381 °C for metal-ceramic fuel (UO2+Al). The calculations show a significant effect of the type of fuel material (UO2 or UO2+Al dispersion composition) on the fuel center temperature, taking into account the operating conditions in the subcritical assembly and the WWR-M research reactor. The maximum temperature of the cladding for the WWR-M operating conditions was 86.5 °C, and the maximum temperature of the cladding for the SCA operating conditions is 27 °C, which does not exceed the boiling point (vaporization) under the nominal conditions of their operation. Cross-section area of fuel rods, heat transfer coefficient and temperature distribution of the coolant are calculated. The software module allowed to estimate the temperature distribution of fuel element with different types of nuclear fuel for the conditions of research nuclear assemblies.

scholarly journals Analysis and comparison of computer programs to analyze the irradiation performance of Uranium Molybdenum monolithic fuel plates and Uranium dioxide cylindrical fuel rods in power reactors.

2021 ◽  
Vol 9 (1) ◽  
Author(s):  
André Luiz Candido da Silva ◽  
Antonio Teixeira e Silva
Keyword(s):  
Uranium Dioxide ◽  
Nuclear Power ◽  
Reactor Core ◽  
Mechanical Analysis ◽  
Operating Conditions ◽  
Nuclear Power Reactor ◽  
Fuel Rods ◽  
Light Water ◽  
Higher Power ◽  
Irradiation Performance

The aim of this work is to present a comparative analysis in terms of the irradiation performance of cylindrical uranium dioxide fuel rods and monolithic uranium molybdenum fuel plates in pressurized light water reactors.To analyze the irradiation performance of monolithic uranium molybdenum fuel plates when subjected to steady state operating conditions in light water pressurized reactors, the computer program PADPLAC-UMo was used, which performs thermal and mechanical analysis of the fuel taking into account the physical , chemicals and irradiation effects to which this fuel is subjected. For the analysis of the uranium dioxide fuel rods, the code FRAPCON was used, which is an analytical tool that verifies the irradiation performance of fuel rods of pressurized light water reactor, when the power variations and the boundary conditions are slow enough for the term permanent regime to be applied. The analysis for a small nuclear power reactor, despite the higher power density applied to the fuel plate in relation to the fuel rod, showed that the fuel plates have lower temperatures and lower fission gas releases throughout the analyzed power history, allowing the use of a more compact reactor core without exceeding the design limits imposed on nuclear fuel.

Axial Fission Gas Transport in Nuclear Fuel Rods

2009 ◽  
Vol 283-286 ◽  
pp. 262-267
Author(s):  
M.T. del Barrio ◽  
Luisen E. Herranz
Keyword(s):  
Nuclear Fuel ◽  
Gas Transport ◽  
Fuel Pellet ◽  
Gas Release ◽  
Gas Mixing ◽  
Fuel Performance ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fission Gas ◽  
Fission Gas Release

Fission of fissile uranium or plutonium nucleus in nuclear fuel results in fission products. A small fraction of them are volatile and can migrate under the effect of concentration gradients to the grain boundaries of the fuel pellet. Eventually, some fission gases are released to the rod void volumes by a thermally activated process. Local transients of power generation could distort even further the already non-uniform axial power and fission gas concentration profiles in fuel rods. Most of the current fuel rod performance codes neglects these gradients and the resulting axial fission gas transport (i.e., gas mixing is considered instantaneous). Experimental evidences, however, highlight axial gas mixing as a real time-dependent process. The thermal feedback between fission gas release, gap composition and fuel temperature, make the “prompt mixing assumption” in fuel performance codes a key point to investigate due to its potential safety implications. This paper discusses the possible scenarios where axial transport can become significant. Once the scenarios are well characterized, the available database is explored and the reported models are reviewed to highlight their major advantages and shortcomings. The convection-diffusion approach is adopted to simulate the axial transport by decoupling both motion mechanisms (i.e., convection transport assumed to be instantaneous) and a stand-alone code has been developed. By using this code together with FRAPCON-3, a prospective calculation of the potential impact of axial mixing is conducted. The results show that under specific but feasible conditions, the assumption of “prompt axial mixing” could result in temperature underestimates for long periods of time. Given the coupling between fuel rod thermal state and fission gas release to the gap, fuel performance codes predictions could deviate non-conservatively. This work is framed within the CSN-CIEMAT agreement on “Thermo-Mechanical Behaviour of the Nuclear Fuel at High Burnup”.

RBMK Fuel Rod Modelling and Investigation of Process During Operation Cycle

2010 ◽  
Author(s):  
Tadas Kaliatka ◽  
Ausˇra Marao ◽  
Renatas Karalevicˇius ◽  
Eugenijus Usˇpuras
Keyword(s):  
Normal Operation ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
The Status ◽  
Fuel Storage ◽  
Spent Fuel Pool ◽  
Operation Cycle ◽  
Rbmk Fuel

This paper presents the analysis determining the status of fuel rods after whole normal operation. The FEMAXI–6 code was selected for such analysis. Evaluating the specifics of RBMK fuel rods, the adaptation of code was provided. After the adaptation of FEMAXI-6 code, the single fuel rod model of RBMK-1500 was developed and the processes, which occur during whole life of fuel rods, were analyzed. For this analysis the fuel rod from fuel channel with average initial power (2.5 MW) was selected. After (normal) operation the fuel rods from the reactor are transferred to the spent fuel pool and the state of the fuel rods (intactness of cladding, residual stresses in the cladding and fuel pellets, gap between cladding and pellets and etc.) is very important, because fuel rod cladding is one of the safety barriers. In this paper the stresses in cladding, plastic deformation of cladding and other parameters were calculated using FEMAXI-6 and method of final elements. The performed analysis demonstrates possibility to identify state of fuel rods after normal operation that is necessary for long-term fuel storage in spent fuel pools.

Improvement of SAMPSON Code for BWR Severe Accident Analysis

2013 ◽  
Author(s):  
Maolong Liu ◽  
Yuki Ishiwatari ◽  
Koji Okamoto
Keyword(s):  
Stainless Steel ◽  
Oxidation Reaction ◽  
Reactor Core ◽  
Severe Accident ◽  
Accident Analysis ◽  
Light Water Reactors ◽  
Fuel Cladding ◽  
Fuel Rod ◽  
Water Reactors ◽  
The Impact

The SAMPSON code has been developed in the IMPACT project in Japan to investigate severe accident phenomena for light water reactors. It integrates various analysis modules into a single code. The authors improved the fuel rod heat-up module of SAMPSON code by modeling the oxidation reaction of various core structures, including Zircaloy, stainless steel and B4C. And the creep failures of the Zircaloy fuel cladding and stainless steel monitoring guide tubes of the source range monitor (SRM) in the reactor core was also modeled for severe accident analysis.

scholarly journals Engineering Margin in Calculations of Energy Release in VVER-1000 core

2018 ◽  
pp. 20-26
Author(s):  
A.M. Abdullayev ◽  
A.I. Zhukov ◽  
S.V. Maryokhin ◽  
S.D. Riabchykov
Keyword(s):  
Energy Release ◽  
Three Dimensional ◽  
Measurement Data ◽  
Reactor Core ◽  
Small Scale ◽  
Two Dimensional ◽  
Fuel Rods ◽  
Fuel Rod ◽  
The Core ◽  
Fuel Assemblies

A method for calculating the engineering margin factor (EMF) in calculations of the energy release in the core of VVER-1000 reactors is proposed in the paper. The analysis of various approaches in the calculation of EMF is carried out and various factors influencing EMF and the ways of their consideration —deterministic and statistical — are determined. The main attention is paid to the influence of gaps between the fuel assemblies on the energy release of fuel rods and the contribution of this factor to the EMF. The limitations and conservatism of two-dimensional small-scale calculations of the energy release of fuel rods in case of deviation of the gap size between the fuel assemblies from the design one are shown. A three-dimensional approach to calculating the contribution of gaps to the EMF is proposed. The approach is based on detailed measurements of the shape of fuel assemblies removed from the core performed at Zaporizhzhya NPP [13]; simulation of the distribution of gaps in the reactor core [16] using measurement data; two-dimensional calculations of the energy release of fuel rods in separate fuel assemblies, surrounded by gaps of different widths, with mirroring boundary conditions; three-dimensional calculations of energy release of fuel rods in fuel assemblies in the reactor core. Two-dimensional and three-dimensional calculations are performed by the wellknown ALPHA-H/PHOENIX-H/ANC-H codes. The proposed approach allows considering not only the change in the fuel rod power, particularly of the peripheral rods, which is inherent in the currently used methods of calculating EMF, but also takes into account the change in the power of the fuel assemblies in the core, which makes the proposed method more realistic and removes the excessive conservatism of EMF calculations and, thereby, allows improving fuel efficiency. For fuel assemblies produced by Westinghouse, it is proposed to use full EMF: for fuel rod power (FΔH) 1.111 and for fuel rod linear power (FQ) 1.173. The use of the BEACONTM monitoring system makes it possible to further reduce the EMF: for fuel rod power (FΔH) - up to 1.084 and for fuel rod linear power (FQ) - up to 1.121.

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