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.

Measurement of Two-Dimensional Void Fraction Distributions of Rising Bubbles in a Simulated Sub-Channel by Wire-Mesh Sensors at Conditions of Forced Convective and Stagnant Flows

2017 ◽  
Author(s):  
Yota Suzuki ◽  
Yusei Tanaka ◽  
Taku Sakka ◽  
Akinori Sato ◽  
Kazuyuki Takase ◽  
...  
Keyword(s):  
Flow Condition ◽  
Two Phase Flow ◽  
Flow Behavior ◽  
Wire Mesh ◽  
Phase Flow ◽  
Two Dimensional ◽  
Two Phase ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Stagnant Flow

Clarifying thermal-hydraulic characteristics in a nuclear reactor core is important in particular to enhance the thermo-fluid safety of nuclear reactors. Spacers installed in subchannels of fuel assemblies have the role of keeping the interval between adjacent fuel rods constantly. Similarly, in case of PWR the spacer has also the role as the turbulence promoter. When the transient event occurs, two-phase flow is generated by boiling of water due to heating of fuel rods. Therefore, it is important to confirm the two-phase flow behavior around the spacer. So, the effect of the spacer affecting the two-phase flow was investigated experimentally at forced convective flow condition. Furthermore, in order to improve the thermal safety of current light water reactors, it is necessary to clarify the two-phase flow behavior in the subchannels at the stagnant flow condition. So, the bubbly flow data around a simulated fuel rod were obtained experimentally at the stagnant flow condition. A wire-mesh sensor was used to obtain a detailed two-dimensional void fraction distribution around the simulated spacer and fuel rod. As a result of this research, the bubbly behavior around the simulated spacer and fuel rod was qualitatively revealed and also bubble dynamics in the sub-channels at the conditions of forced convective and stagnant flows were evaluated. The present experimental data are very useful for verifying the detailed three-dimensional two-phase flow analysis codes.

scholarly journals Lumped parameter modelling of two-phase ejectors: numerical implications of the equilibrium assumptions

2021 ◽  
Vol 2116 (1) ◽  
pp. 012089
Author(s):  
Giorgio Besagni
Keyword(s):  
Carbon Dioxide ◽  
Critical Temperature ◽  
Operating Conditions ◽  
Cutting Edge ◽  
Technical Solution ◽  
Two Phase ◽  
Operation Conditions ◽  
System A ◽  
Lumped Parameter ◽  
Major Shortcoming

Abstract The use of carbon dioxide as refrigerant is attracting a growing attention and is a cutting-edge research topic. In spite of its many advantages, carbon dioxide has a major shortcoming, viz., low critical temperature. Owing to the low critical temperature, carbon dioxide cycles encompass both the sub-critical and the trans-critical operation conditions; the trans-critical operating conditions are characterized by high thermodynamic losses, requiring particular attention in the integrated component/system design criteria. In this perspective, in recent years, ejector technology has been widely recognized as a promising technical solution to support the deployment of carbon dioxide cycles, by reducing throttling losses. Unfortunately, the large variation in system operations as well as the changes in sub-critical and trans-critical operating conditions makes the numerical simulation of carbon dioxide ejector-based system a cutting-edge challenge. This paper contributes to the present day discussion on the validation of lumped parameter models for carbon dioxide ejectors. A model taken from the literature has been tested against literature data and the equilibrium assumptions, underlying the modelling approach have been tested.

Investigation Into the Influence of Fuel Rods on the Anti-Seismic Mechanical Characteristics of Fuel Assembly Spacer Grid by FEM Method

2017 ◽  
Author(s):  
Jingya Sun ◽  
Yu Dang ◽  
Song Liu ◽  
Jiazheng Liu ◽  
Libing Zhu
Keyword(s):  
Fuel Assembly ◽  
Simulation Method ◽  
Experimental Results ◽  
Spacer Grid ◽  
Fuel Rods ◽  
Fem Model ◽  
Spacer Grids ◽  
Fuel Rod ◽  
Fem Method ◽  
Crush Strength

The anti-seismic performance of fuel assembly is mainly determined by the critical crush load and the stiffness of spacer grids. To comprehensive know about the influence of fuel rods on the spacer grid, a 5×5 spacer grid FEM model which including fuel rods is established. Basing the fact that the grid spring has remarkable influence on the grid crush strength which is found in experiment, some cases are carried out, which are used to analyze effects of grid with/without fuel rod, friction between the grid spring/dimple and the fuel rod, the deflection of grid spring on the static buckling strength. Results show that grids with fuel rods will have higher crush strength than those without fuel rods; at certain range, increasing grid spring deflection at working point will do help to increase the grid crush strength; higher friction coefficient of grid spring and fuel rod can enhance the crush strength. Comparing with experimental results in literatures, results from simulations show the same tendency with the experimental results. The conclusion and the simulation method involved in this paper can provide some guidelines to optimize the performance of spacer grid assembly.

Flow-Induced Grid-to-Rod Fretting Test and Simulation for NFI PWR Fuel Assembly

2013 ◽  
Author(s):  
Shota Okui ◽  
Yuichiro Kubo ◽  
Shumpei Kakinoki ◽  
Roger Y. Lu ◽  
Zeses Karoutas ◽  
...  
Keyword(s):  
Numerical Simulation ◽  
Fuel Assembly ◽  
Wear Test ◽  
Flow Induced Vibration ◽  
Fuel Rods ◽  
Spacer Grids ◽  
Fuel Rod ◽  
Non Linear ◽  
Computational Fluid Dynamics Cfd

A long-term flow-induced vibration and wear test was performed for a full-scale 17×17 PWR fuel mockup, and the test results were compared with numerical simulations. The flow-induced vibration on a fuel assembly or fuel rods may cause Grid-to-Rod Fretting (GTRF) and result in the leakage of fuel rods in PWRs. GTRF involves non-linear vibration of a fuel rod due to the excitation force induced by coolant flow around a fuel rod. So, the numerical simulation is performed by VITRAN (Vibration Transient Analysis Non-linear) and Computational Fluid Dynamics (CFD). VITRAN code was developed by Westinghouse to simulate fuel rod flow induced vibration and GTRF. In this paper, it was confirmed that the code can reproduce GTRF wear for NFI fuel assembly. CFD calculation is performed to obtain the axial and lateral flow velocity around the fuel rods, reflecting detailed geometries of fuel assembly components like bottom nozzle, spacer grids. The numerical simulation reasonably reproduced the vibration and wear test for NFI fuel assembly.

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.

Three Dimensional Modeling of the Thermo-Mechanical Performance of the Fuel Rods of a PWR

2017 ◽  
Author(s):  
Wang Zhu ◽  
Zhang Chunyu ◽  
Li Aolin ◽  
Yuan Cenxi
Keyword(s):  
Mechanical Performance ◽  
Three Dimensional ◽  
Fission Products ◽  
General Purpose ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Three Dimensional Modeling ◽  
Water Reactors

The fuel rods of pressurized water reactors operate under complex radioactive, thermal and mechanical conditions. Multiphysics has to be taken into account in order to evaluate their performance. Many existing fuel rod codes make great simplifications on analyzing the behavior of fuel rods. The present study develops a three dimensional module within the framework of a general-purpose finite element solver, i.e. ABAQUS, for modeling the thermo-mechanical performance of the fuel rods. A typical fuel rod is modeled and the temperature as well as the stress within the pellets are computed. The results show that the burnup levels have an important influence on the fuel temperature. The swelling of fission products cause dramatically increasing of pellet strain. The change of the cladding stress and radial displacement with the axial length can be reasonably predicted. It is shown that a quick power ramp or a reactivity insertion accident can induce high tensile stress to the outer regime of the pellet and may cause further fragmentation to the pellets.

Development of a Multiphysics Approach With ANSYS CFX and PARCS for Local Analysis of Neutron Oscillation

2017 ◽  
Author(s):  
Zhuoqi Du ◽  
Marcus Seidl ◽  
Rafael Macián-Juan
Keyword(s):  
Fuel Assembly ◽  
Power Distribution ◽  
Noise Analysis ◽  
Flow Distribution ◽  
Local Analysis ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Ansys Cfx ◽  
Neutron Noise

Neutron noise analysis has been done over the decades to predict fuel assembly vibrations and to evaluate safety related issues. Neutron noise occurs due to several reasons: the vibration of the fuel rods, flow obstacles such as rod bending and crud deposition, the moderator temperature and time dependent changes caused by varying flow distributions within a fuel assembly, etc. In order to have a better insight of the neutron noise, a fluid mechanics, structural and neutronics coupled code is developed to perform detailed multiphysics simulations at the level of the fuel rods inside a fuel assembly. In this paper the coupling routine of both steady state and transient calculation is described and the outcome is discussed under several scenarios to understand the influence of rod vibration, moderator temperature and flow distribution on the neutronic field. This paper presents the methodology to couple the multiphysics Computational Fluid Dynamics (CFD) code ANSYS-CFX 16.0 with the 3D neutron diffusion code PARCS v3.0. The model for a 16×16 Pressurized Water Reactor (PWR) fuel assembly is set up for ANSYS-CFX. A sensitivity analysis is carried out to obtain the optimal mesh parameters which results in a good accuracy, as well as a small need for computation capability. Transient cases are studied on a quarter fuel assembly applying oscillating moderator inlet boundary conditions in which the inlet moderator temperature and the inlet moderator velocity are varying over time. In order to simulate the vibration of the fuel rod, the fuel rod part is implemented as immersed solid in ANSYS-CFX. Different vibration modes are applied to both cases: individual single rods of the fuel assembly, and all rods of the fuel assembly. The results of each case are shown in this paper giving a better understanding of how axial power distribution develops with varying flow conditions and vibrating fuel rods.

Research on Manufacturing Technology of China Experimental Fast Reactor Fuel Rod

2016 ◽  
Author(s):  
Yu Hu ◽  
Jianhong Wang
Keyword(s):  
Fuel Assembly ◽  
Fast Reactor ◽  
Nuclear Industry ◽  
Reactor Fuel ◽  
Fuel Rods ◽  
Process Step ◽  
Research Technology ◽  
Fuel Rod ◽  
Fast Reactor Fuel ◽  
China Experimental Fast Reactor

CEFR, which is the acronym of China Experimental Fast Reactor, as the fourth generation advanced nuclear power system, plays a very important role in the development of nuclear industry in the future. Currently, as a kind of study, has not yet formed a fuel rod and fuel assembly manufacturing company in China. Realizing the localization of fast reactor fuel rod, fuel assembly and equipment has a very positive meaning to China’s nuclear industry development. At the same time for the CJNF’s development, promoting the manufacture level and scientific research technology of product, the research plays an extremely important role. In this paper, a fuel rod extrusion pit device and a fuel rod cladding pipe wire equipment have been designed according to the characteristic of fast reactor components. The fuel rod extrusion pit device to ensure pit deepness is 1.8–2.0mm, the radius of pit is 3mm, three pits into each other 120°, deviation is ±5°, the distance is 450.4±0.3mm. And the fuel rod cladding pipe wire equipment meet the product technical requirements for the pitch of 100±5mm, each distance deviation shall not be more than ±15mm, wire tension force for 100±20N. Massive tests on the extrude pits, press plug, load pellet, wire and spot welding, fuel rods ring welding and other important process step have been conducted and develop a set of optimized production plan. Fast reactor fuel rods structure have a enormous difference with the past other types of products have, the control of the input power for welding and an increase in the pressure of sealed welding chambers have been employed and avoided the problem of weld inflatable and weld porosity successfully. Results showed that the study on manufacturing of fast reactor fuel rod achieves success and realize the localization of fast reactor fuel rod in CJNF.

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.

Study on the Thermal-Mechanical Performance of SiC Composites Cladding Under Multiple Conditions

2020 ◽  
Author(s):  
Chunyu Yin
Keyword(s):  
Failure Probability ◽  
Hoop Stress ◽  
Mechanical Performance ◽  
Performance Model ◽  
Operating Conditions ◽  
Analysis Tool ◽  
Operation Conditions ◽  
Sic Fiber ◽  
Maximum Hoop Stress ◽  
Sic Composites

Abstract SiC has become a candidate cladding material of Accident Tolerant Fuels (ATF) due to its excellent irradiation stability and corrosion resistance. However, because SiC is a ceramic material with low toughness, brittle failure is a significant concern. In order to improve the toughness, SiC fiber is required to manufacture multi-layer SiC composites. But the current performance model or analysis tool is not available for SiC composites cladding due to its obviously difference with Zr alloy cladding. On one side, Finite element method was used in this paper to analyze the performance of SiC composites cladding under operation conditions which include normal, transient conditions and LOCA conditions; on the other side, this paper gives the performance of the SiC composites with two layers under multiple operating conditions. The result showed that the temperature was stable and the maximum hoop stress was reached at about 70d under normal condition. The power ramp can increase the cladding temperature and has visible influence on the stress distribution. The hoop stress of the cladding reversed under LOCA condition. The tensile hoop stress on the outer surface significantly increased, which caused the obvious increase of failure probability of monolithic SiC, and the failure probability of SiC layer is significantly increased. The conclusion of the analysis has guiding significance for the theoretical design of SiC composites.

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