Modelling and simulation of the primary system for a small lead-cooled fast reactor with a ratio of core power to flow

2021 ◽  
pp. 108829
Author(s):  
Yang Hu ◽  
Lehua Liang ◽  
Chuhao Li ◽  
Xiaoyu Li ◽  
Wenjie Zeng ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Modelling And Simulation ◽  
Primary System

Thermal-Hydraulic Simulations of Guard Containment in Depressurization Accidents of a Gas-Cooled Fast Reactor

2013 ◽  
Author(s):  
Zhanjie Xu ◽  
Thomas Jordan
Keyword(s):  
Fast Reactor ◽  
Nuclear Reactor ◽  
Source Term ◽  
Reference Value ◽  
Design Stage ◽  
Primary System ◽  
Simulation Results ◽  
Accident Scenarios ◽  
Design Pressure ◽  
The Eu

A gas-cooled fast reactor is designed as an advanced nuclear reactor in next generation in the EU. In depressurization accident scenarios, pressurization caused by a release of helium from the primary system with a higher pressure into the guard containment would endanger the integrity of the containment. In the design stage, the released source term is analyzed theoretically, and is applied as a boundary condition in the 3D CFD code simulation to the transient pressurization process. The simulation results supply a reference value about the design pressure of the containment.

Implementation of Liquid Metal Properties in RELAP5 MOD3.2 for Safety Analysis of Sodium-Cooled Fast Reactors

2017 ◽  
Author(s):  
Jian Song ◽  
Limin Liu ◽  
Simiao Tang ◽  
Yingwei Wu ◽  
Wenxi Tian ◽  
...  
Keyword(s):  
Experimental Data ◽  
Liquid Metal ◽  
Fast Reactor ◽  
Natural Circulation ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Working Fluid ◽  
Primary System ◽  
Steady State Condition ◽  
Design Values

Due to great deal of operation experience and technology accumulation, sodium cooled fast reactor (SFR) is the most promising among the six Generation IV reactors, which has advantages of breeding nuclear fuel, transmuting long-lived actinides and good safety characteristics. Thermal-hydraulic computer codes will have to be developed, verified, and validated to support the conceptual and final designs of new SFRs. However, work on developing thermal hydraulic analysis code for SFR is very limited in China, while the common software RELAP5 MOD3 is unable to analyze liquid metal systems. So the modified RELAP5 MOD3.2 is being considered as the thermal-hydraulic system code to support the development of the SFRs. The thermodynamic and transport properties of sodium liquid and vapor have been implemented into the RELAP5 MOD3.2 code, as well as the specific heat transfer correlations for liquid metal. The sodium liquid properties use polynomial equations based on data obtained from Argonne National Laboratory, and the vapor is assumed to be perfect gas. The property equations are acceptably accurate for analysis of SFR, especially for single-phase liquid. New files are added to the fluids directory to generate property tables for new working fluid, which are similar to the table interpolation subroutines for light and heavy water in the original file directory. The method of code modifications are universal for other working fluids and will not affect the code original performance. Some basic verification work for the modified code are carried out. The steam generator of CEFR is analyzed to verify the modified code. The calculated results show that all the water will boil off in the evaporator and the calculated results are in good agreement with the design values. By using modified RELAP5 to model the primary loop of EBR-II fast reactor, the SHRT-17 PLOF test was analyzed. The results show that the natural circulation can be established in the EBR-II primary system after main pumps off to remove the core decay residual heat effectively, and the peak temperature under the safety limits. Moreover, the results computed in this work compared well with the test experimental data for the steady state condition. During the transients, the changing trends of temperature and pressure are similar to experimental data. The discrepancies between calculation and experiment are considered acceptably which need to be improved in the future work. Our work could demonstrate the capability and reliability of the modified RELAP5 for the analysis of SFRs further.

Modeling of Reactivity Effects and Transient Behaviour of Large Sodium Fast Reactor

2021 ◽  
Author(s):  
Alexander Ponomarev ◽  
Konstantin Mikityuk
Keyword(s):  
Fast Reactor ◽  
Core Model ◽  
Primary System ◽  
Control Rod ◽  
Transient Behaviour ◽  
The Core ◽  
Sodium Fast Reactor ◽  
Reactivity Coefficients ◽  
Core Conditions ◽  
Point Kinetics

Abstract In the paper the reactivity characteristics of the core of the large sodium fast reactor Superphenix (SPX) were evaluated and compared with available experimental data. The analysis was performed using the TRACE system code modified for the fast reactor applications. The simplified core model was developed aiming to overcome the lack of detailed information on design and realistic core conditions. Point Kinetics neutronic model with all relevant reactivity feedbacks was used to calculate transient power. The paper focuses on challenging issue of modelling of the transient thermal responses of primary system structural elements resulting in reactivity feedbacks specific to such large fast reactor which cannot be neglected. For these effects, the model was equipped with dedicated heat structures to reproduce important feedbacks due to vessel wall, diagrid, strongback, control rod drive lines thermal expansion. Peculiarly, application of the model was considered for a whole range of core conditions from zero power to 100% nominal. The developed core model allowed reproducing satisfactorily the core reactivity balance between zero power at 180?C and full power conditions. Additionally, the reactivity coefficients k, g, h at three power levels were calculated and satisfactory agreement with experimental measurements was also observed. The study demonstrated feasibility of application of relatively simple model with adjusted parameters for analysis of different conditions of very complex system.

Performance of Key Features of EBR-II and the Implications for Next Generation Systems

2002 ◽  
Author(s):  
Ronald W. King ◽  
Douglas L. Porter
Keyword(s):  
Fast Reactor ◽  
Fuel Cycle ◽  
The United States ◽  
Reactor Power ◽  
Early Years ◽  
Primary System ◽  
Next Generation ◽  
Successful Operation ◽  
Extensive Evaluation ◽  
Key Features

The Experimental Breeder Reactor No. 2 (EBR-II) began operation in August 1964 as an engineering test of the operation of a sodium-cooled fast reactor power plant system. Its primary mission was to demonstrate and evaluate the performance of a sodium-cooled reactor system using recycled fuel from an integral fuel cycle facility, and as an electrical power generator tied to the utility grid. It accomplished this mission in the early years of operation. Following the early successes, the mission evolved to include more extensive testing of fuels, materials, components, and safety features of a sodium-cooled fast reactor. The use of sodium as the coolant, use of metal fuel, and a piped-pool primary system configuration, were key contributors to its notable long-term performance. Extensive evaluation and examination of these features have provided a solid basis for and understanding of the technology. Recent interest in future designs for nuclear generating stations has generated renewed interest in liquid-metal-cooled fast reactors in the United States and elsewhere. The successful operation of EBR-II for thirty years and the demonstration of characteristics significant to the development of next-generation reactors, has prompted a re-examination of key features of EBR-II and a review of its performance record. This paper discusses selected key features and their contribution to the performance of EBR-II, evaluates the overall performance of the reactor, and discusses the implications for the development of next generation reactor concepts.

Parametric Study of Sloshing Effects in the Primary System of an Isolated Lead-Cooled Fast Reactor

2015 ◽  
Vol 190 (1) ◽  
pp. 1-10 ◽  
Author(s):  
Marti Jeltsov ◽  
Walter Villanueva ◽  
Pavel Kudinov
Keyword(s):  
Fast Reactor ◽  
Parametric Study ◽  
Primary System

Modeling and Control Strategy of the China LBE Cooled Fast Reactor

2014 ◽  
Author(s):  
Shoujun Yan ◽  
Zhao Wang ◽  
Pengfei Wang ◽  
Jiashuang Wan ◽  
Huawei Fang ◽  
...  
Keyword(s):  
Control System ◽  
Dynamic Characteristics ◽  
Fast Reactor ◽  
Control Strategy ◽  
Pi Controller ◽  
Primary System ◽  
Air Cooling ◽  
Feed Forward ◽  
Lead Bismuth Eutectic ◽  
And Control

China lead bismuth eutectic (LBE) cooled fast reactor comprises of the primary system with lead bismuth eutectic (LBE) as the coolant, the secondary circuit with sub-cooled water as the coolant and the associated air cooling system for an effective rejection of thermal power to the environment as a final heat sink. The dynamic characteristics of LBE cooled fast reactor are different from the traditional Pressurized Water Reactors (PWRs) because of the variances in coolant properties as well as major differences due to the operation in the fast versus the thermal neutron spectrum. To investigate the dynamic characteristics of the CLEAR-IA reactor for control system design and simulation, a model for the main components of the reactor and the most relevant interactions among them is developed. Since all the coefficients in the models are functions of temperature, the models in this paper are not described by ordinary differential equation. These models are realized by using the S-function builder of SIMULINK. The steady state calculation result based on the thermal-hydraulic models show agreement with the design value. To show the proposed models could be used for the simulation, the transient process of parameter change is compared with Relap5 simulation code, which shows agreement. A Proportional-Integral (PI) controller is designed to keep the power following the set value as quickly as possible. To keep the inlet temperature of core coolant constant, a control strategy based on a simultaneous feed-forward and feedback scheme has been adopted. The feed-back control system is a PI controller and it can achieve a real time and no error control, but it has time delay. The feed-forward control can realize the control in advance before the LBE temperature at inlet of the core has been changed to reduce the overshoot. So the feed-forward can realize an advance and rough control, the feedback can realize a no error and accurate control. Based on the developed model and control strategy, dynamic simulations of the whole system in case of step changes of reactivity and set power are performed. The simulation results show that the proposed model is accurate enough to describe the dynamic behaviors of the plant in spite of its simplicity. It has also been demonstrated that the developed controllers for the CLEAR-IA can provide superior reactor control due to the efficiency of the control strategy adopted.

Current Status of the Development of the Fast Reactor Fuel Cycle in India

2009 ◽  
pp. 120-126
Author(s):  
K.V. Govindan Kutty ◽  
P.R. Vasudeva Rao ◽  
Baldev Raj
Keyword(s):  
Fast Reactor ◽  
Fuel Cycle ◽  
Reactor Fuel ◽  
Current Status ◽  
Fast Reactor Fuel
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