A coupled analysis of system thermal-hydraulics and three-dimensional reactor kinetics for a 12-finger control element assembly drop event in a PWR plant

2010 ◽  
Vol 37 (11) ◽  
pp. 1580-1587 ◽  
Author(s):  
Jae Jun Jeong ◽  
Seung Wook Lee ◽  
Jin Young Cho ◽  
Bub Dong Chung ◽  
Gyu-Cheon Lee
Keyword(s):  
Three Dimensional ◽  
Coupled Analysis ◽  
Control Element ◽  
Thermal Hydraulics ◽  
Reactor Kinetics ◽  
Finger Control

scholarly journals Neutronic and thermal-hydraulic coupling for 3D reactor core modeling combining MCB and fluent

Nukleonika ◽  
2015 ◽  
Vol 60 (3) ◽  
pp. 531-536 ◽  
Author(s):  
Igor P. Królikowski ◽  
Jerzy Cetnar
Keyword(s):  
Cross Sections ◽  
Nuclear Reactions ◽  
Nuclear Reactors ◽  
Input Parameter ◽  
Three Dimensional ◽  
Reactor Core ◽  
Coupled Analysis ◽  
Thermal Hydraulics ◽  
Power Profile ◽  
The Impact

Abstract Three-dimensional simulations of neutronics and thermal hydraulics of nuclear reactors are a tool used to design nuclear reactors. The coupling of MCB and FLUENT is presented, MCB allows to simulate neutronics, whereas FLUENT is computational fluid dynamics (CFD) code. The main purpose of the coupling is to exchange data such as temperature and power profile between both codes. Temperature required as an input parameter for neutronics is significant since cross sections of nuclear reactions depend on temperature. Temperature may be calculated in thermal hydraulics, but this analysis needs as an input the power profile, which is a result from neutronic simulations. Exchange of data between both analyses is required to solve this problem. The coupling is a better solution compared to the assumption of estimated values of the temperatures or the power profiles; therefore the coupled analysis was created. This analysis includes single transient neutronic simulation and several steady-state thermal simulations. The power profile is generated in defined points in time during the neutronic simulation for the thermal analysis to calculate temperature. The coupled simulation gives information about thermal behavior of the reactor, nuclear reactions in the core, and the fuel evolution in time. Results show that there is strong influence of neutronics on thermal hydraulics. This impact is stronger than the impact of thermal hydraulics on neutronics. Influence of the coupling on temperature and neutron multiplication factor is presented. The analysis has been performed for the ELECTRA reactor, which is lead-cooled fast reactor concept, where the coolant fl ow is generated only by natural convection

Coupled Analysis of Thermal Hydraulics and Neutronics for a Molten Salt Reactor

2017 ◽  
Author(s):  
Jian Ge ◽  
Dalin Zhang ◽  
Wenxi Tian ◽  
Suizheng Qiu ◽  
G. H. Su
Keyword(s):  
Molten Salt ◽  
Nuclear Reactor ◽  
Coupled Analysis ◽  
Molten Salt Reactor ◽  
Thermal Hydraulics ◽  
Delayed Neutron ◽  
Analysis Model ◽  
Liquid Salt ◽  
Energy Equations ◽  
Volume Method

As one of the six selected optional innovative nuclear reactor in the generation IV International Forum (GIF), the Molten Salt Reactor (MSR) adopts liquid salt as nuclear fuel and coolant, which makes the characteristics of thermal hydraulics and neutronics strongly intertwined. Coupling analysis of neutronics and thermal hydraulics has received considerable attention in recent years. In this paper, a new coupling method is introduced based on the Finite Volume Method (FVM), which is widely used in the Computational Fluid Dynamics (CFD) methodology. Neutron diffusion equations and delayed neutron precursors balance equations are discretized and solved by the commercial CFD package FLUENT, along with continuity, momentum and energy equations simultaneously. A Temporal And Spatial Neutronics Analysis Model (TASNAM) is developed using the User Defined Functions (UDF) and User Defined Scalar (UDS) in FLUENT. A neutronics benchmark is adopted to demonstrate the solution capability for neutronics problems using the method above. Furthermore, a steady state coupled analysis of neutronics and thermal hydraulics for the Molten Salt Advanced Reactor Transmuter (MOSART) is performed. Two groups of neutrons and six groups of delayed neutron precursors are adopted. Distributions of the liquid salt velocity, temperature, neutron flux and delayed neutron precursors in the core are obtained and analyzed. This work can provide some valuable information for the design and research of MSRs.

Coupled Three-Dimensional Neutronics and Thermal-Hydraulics Analysis for SCWR Core Typical Transients

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Wang Lianjie ◽  
Lu Di ◽  
Zhao Wenbo
Keyword(s):  
Electric Power ◽  
Supercritical Water ◽  
Transient Analysis ◽  
Three Dimensional ◽  
Thermal Hydraulics ◽  
Water Density ◽  
Fuel Temperature ◽  
Control Rod ◽  
Reactor Protection System ◽  
Supercritical Water Cooled Reactor

Transient performance of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWel (CSR1000) core during some typical transients, such as control rod (CR) ejection and uncontrolled CR withdrawal, is analyzed and evaluated with the coupled three-dimensional neutronics and thermal-hydraulics SCWR transient analysis code. The 3D transient analysis shows that the maximum cladding surface temperature (MCST) retains lower than safety criteria 1260 °C during the process of CR ejection accident, and the MCST retains lower than safety criteria 850 °C during the process of uncontrolled CR withdrawal transient. The safety of CSR1000 core can be ensured during the typical transients under the salient fuel temperature and water density reactivity feedback and the essential reactor protection system.

Three-dimensional thermal and electromagnetic coupled analysis of railguns

1993 ◽  
Vol 29 (1) ◽  
pp. 356-361 ◽  
Author(s):  
B. Azzerboni ◽  
G. Tina ◽  
E. Cardelli ◽  
M. Raugi
Keyword(s):  
Three Dimensional ◽  
Coupled Analysis

Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

2006 ◽  
Author(s):  
A. Gorzel
Keyword(s):  
Three Dimensional ◽  
Reactor Core ◽  
Reactor Power ◽  
Pressure Increase ◽  
Thermal Hydraulics ◽  
Short Term ◽  
Speed Reduction ◽  
Pump Speed ◽  
Transient Behaviour ◽  
Limit Value

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and impermissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a German BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second — much smaller — maximum that would occur around one second after the first one in the absence of a SCRAM.

Neutronics and Thermal Hydraulics Coupled 3-D Calculation of Pressure-Tube Deformation

2020 ◽  
Author(s):  
Eunhyun Ryu ◽  
Hangyu Joo ◽  
Seungyul Yoo ◽  
Jongyub Jung
Keyword(s):  
Dimensional Space ◽  
Three Dimensional ◽  
Safety Margin ◽  
Pressure Tube ◽  
Coolant Temperature ◽  
Main Concern ◽  
Thermal Hydraulics ◽  
Pressure Tubes ◽  
Fuel Pin ◽  
Heavy Water Reactor

Abstract Among the various parts in a pressurized heavy-water reactor (PHWR), pressure tubes are of tremendous importance. This is because they withstand extreme both pressure and temperature differences that exist between the Primary Heat Transport System (PHTS) and the moderator. The pressure tubes also contribute to prevention of fission product release from the PHTS to the PHWR plant (together with end fittings and nearby parts including plugs). When a PHWR is given a 1% derating, half is due to the aging of the pressure tubes. The main concern with pressure tubes is decrease of the safety margin. Most of the reduction comes from the effects caused by radial expansion and axial sagging, which are belong to four major phenomena including the thinning and the elongation. More specifically, the fuel-pin temperature distribution changes for the worse if deformation of the pressure tube occurs. Because there is extreme irradiation inside the core, the tube content is exposed to high temperature and high pressure. Thus, the shape of the pressure tube is deformed as times goes on. In this paper, using modeling of a deformed pressure tube in three-dimensional space, the effects on the fuel, coolant temperature, and coolant density, were studied quantitatively. This included a neutronics effect explored using coupled neutronics and thermal hydraulics (T/H) calculations. Among the results, only marginal changes of the neutronics effects were observed. The T/H results, which included temperature and density of the fuel and the coolant, were not critical. Through this study, we are now able to determine in new ways, conventional derating values from a pressure tube.

Realization of Uniform Deformation under Three-Dimensional Condition Based on Soil-Water Coupled Analysis Considering Inertia Forces

2014 ◽  
Vol 638-640 ◽  
pp. 530-533
Author(s):  
Bin Bin Xu ◽  
Toshihiro Noda ◽  
Kentaro Nakai
Keyword(s):  
Soil Water ◽  
Three Dimensional ◽  
Coupled Analysis ◽  
Deformation Analysis ◽  
Localized Deformation ◽  
Uniform Deformation ◽  
Soil Skeleton ◽  
Theoretical Considerations ◽  
Inertia Forces ◽  
Dimensional Condition

In the paper, based on soil-water coupled finite deformation analysis, theoretical considerations and numerical calculations were carried out under undrained three-dimensional condition in order to reproduce a uniform deformation field. At first, a theoretical consideration was assumed to realize a uniform deformation for a saturated soil, according to which the initial velocity and acceleration in both vertical and circumferential directions should be applied to each node to remove the influence of inertia effect. This first theoretical analysis is useful and can guide the numerical calculation. Next, the paper realized a uniform deformation of a three-dimensional cylinder specimen under undrained boundary conditions using the soil-water coupled analysis in which the SYS Cam-clay model is employed as the constitutive model for soil skeleton. The numerical results show that without the inertia forces there is no localized deformation in the specimen.

Three-Dimensional Finite Element Analysis of Skin-Pass Rolling and New Models for Process Control

2017 ◽  
Vol 139 (9) ◽  
Author(s):  
Sung Jin Yoon ◽  
Tae Jin Shin ◽  
Jae Sang Lee ◽  
Sang Moo Hwang
Keyword(s):  
Finite Element Analysis ◽  
Finite Element ◽  
Three Dimensional ◽  
Coupled Analysis ◽  
Stress Profile ◽  
Fe Model ◽  
Element Analysis ◽  
Skin Pass ◽  
Dimensional Finite Element ◽  
Three Dimensional Finite Element

This paper describes in detail the deformation behavior of the rolls and strip predicted from the three-dimensional finite element analysis of skin-pass rolling. The predictions are made on the basis of the coupled analysis of elastic deformation of the rolls and elastic–plastic deformation of the strip. Predictions from the proposed finite element (FE) model are compared with experimental data from laboratory-scale cold rolling mills. Then, proposed are models for the prediction of the roll force profile and for the prediction of the residual stress profile. The prediction accuracy of the models is examined through comparison with the predictions from the FE model.

Safety Analysis of the Control Rod Drop Event Applying 3-D Neutronics/Thermal-Hydraulics Coupling Code SPARKLE

2009 ◽  
Author(s):  
Yuta Maruyama ◽  
Satoshi Imura ◽  
Junto Ogawa ◽  
Shuhei Miyake
Keyword(s):  
Power Distribution ◽  
Transient Analysis ◽  
Three Dimensional ◽  
Safety Margin ◽  
Safety Analysis ◽  
Thermal Hydraulics ◽  
Analysis Method ◽  
Dynamic Changes ◽  
Control Rod ◽  
Ejection Event

Mitsubishi Heavy Industries (MHI) has developed the SPARKLE code, which is a PWR plant system transient analysis code that includes a three-dimensional (3D) neutronics module coupled with a thermal-hydraulics module. MHI has performed a study of the applicability of the SPARKLE code to the events which are associated with dynamic changes in power distribution, such as the rod ejection event or the steam line break event. In this paper, MHI has applied the SPARKLE code to the control rod drop event (drop of multiple rods), which features such a power distribution change. In addition, the neutron flux detection is dependent on the location of the dropped rods in this event, which can be dynamically calculated in the SPARKLE code. By applying the SPARKLE code to the control rod drop event, it was confirmed that the safety margin for this event is sufficiently larger than the margin calculated using the current safety analysis method, even if the appropriate conservative assumptions are made.

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