CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Jing Chen ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Kui Zhang ◽  
Mingjun Wang ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Three Dimensional ◽  
Reactor Core ◽  
Heat Removal ◽  
The Core ◽  
Computational Fluid Dynamics Cfd ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Full Core

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.

scholarly journals Analysis on passive residual heat removal system with heat pipes for longterm decay heat removal of small lead-based reactor

2021 ◽  
Vol 236 ◽  
pp. 01018
Author(s):  
Chongju Hu ◽  
Wangli Huang ◽  
Zhizhong Jiang ◽  
Qunying Huang ◽  
Yunqing Bai ◽  
...  
Keyword(s):  
Three Dimensional ◽  
Heat Removal ◽  
Heat Pipes ◽  
Decay Heat ◽  
The Core ◽  
Accident Condition ◽  
Computational Fluid Dynamics Cfd ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

.A lead-based reactor with employing heat pipes as passive residual heat removal system (PRHRS) for longterm decay heat removal was designed. Three-dimensional computational fluid dynamics (CFD) software FLUENT was adopted to simulate the thermal-hydraulic characteristics of the PRHRS under Station-Black-Out (SBO) accident condition. The results showed that heat in the core could be removed smoothly by the PRHRS, and the core temperature difference is less than 20 K.

Calculation studies of coated particles performance in sodium-cooled fast reactor

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 45-49
Author(s):  
N. V. Maslov ◽  
E. I. Grishanin ◽  
P. N. Alekseev
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Heat Removal ◽  
Design Solution ◽  
Fast Neutrons ◽  
Steel Shell ◽  
Primary Circuit ◽  
Coated Particles ◽  
The Core ◽  
Fuel Assemblies

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

scholarly journals CFD Analysis of the Passive Decay Heat Removal System of an LBE-Cooled Fast Reactor

2018 ◽  
Vol 2018 ◽  
pp. 1-11
Author(s):  
Jiarun Mao ◽  
Lei Song ◽  
Yuhao Liu ◽  
Jiming Lin ◽  
Shanfang Huang ◽  
...  
Keyword(s):  
Heat Exchange ◽  
Fast Reactor ◽  
Natural Circulation ◽  
Heat Removal ◽  
Engineering Project ◽  
Decay Heat ◽  
The Core ◽  
Main Vessel ◽  
Heat Removal System ◽  
Removal System

This paper presents capacity of the passive decay heat removal system (DHRS) operated under the natural circulation conditions to remove decay heat inside the main vessel of the Lead-bismuth eutectic cooled Fast Reactor (LFR). The motivation of this research is to improve the inherent safety of the LFR based on the China Accelerator Driven System (ADS) engineering project. Usually the plant is damaged due to the failure of the main pumps and the main heat exchangers under the Station Blackout (SBO). To prevent this accident, we proposed the DHRS based on the diathermic oil cooling for the LFR. The behavior of the DHRS and the plant was simulated using the CFD code STAR CCM+ using LFR with DHRS. The purpose of this analysis is to evaluate the heat exchange capacity of the DHRS and is to provide the reference for structural improvement and experimental design. The results show that the stable natural circulations are established in both the main vessel and the DHRS. During the decay process, the heat exchange power is above the core decay heat power. In addition, in-core decay heat and heat storage inside the main vessel are efficiently removed. All the thermal-hydraulics parameters are within a safe range. Moreover, the highest temperature occurs at the upper surface of the core. A swirl occurs at the corner of the lateral core surface and some improvements should be considered. And the natural circulation driving force can be further increased by reducing the loop resistance or increasing the natural circulation height based on the present design scenario to enhance the heat exchange effect.

Effect of Deformation of Core Elements of Fast Reactor Core to the Seismic Response

2019 ◽  
Author(s):  
Akihisa Iwasaki ◽  
Shinichiro Matsubara ◽  
Kazuteru Kawamura ◽  
Hidenori Harada ◽  
Tomohiko Yamamoto
Keyword(s):  
Thermal Expansion ◽  
Fast Reactor ◽  
Seismic Analysis ◽  
Three Dimensional ◽  
Reactor Core ◽  
Axial Direction ◽  
Vertical Displacement ◽  
Analysis Model ◽  
The Core ◽  
Core Elements

Abstract Core elements of a fast reactor are self-standing on the core support structure and not restrained in the axial direction. When the earthquake occurs, it is necessary to consider vertical behavior and horizontal displacement of the core elements simultaneously. In the core seismic analysis, a three dimensional core vibration behavior was evaluated by considering fluid structure interaction, collision with adjacent core elements and vertical displacement and verified by a series of vibration tests. But the evaluation had a assumption of straightness of each core elements which may be bowed due to thermal expansion and swelling under restraint of the horizontal direction between the upper pad and lower structure (Entrance Nozzle). If the core elements are deformed in its plant operation, they may push each other against its adjacent core elements. The large horizontal interference forces may work to decrease the vertical displacement of the core elements. In this study, to grasp and estimate the behavior under the deformed core elements under the earthquake motion, a three dimensional seismic analysis model consist of all of core elements with consideration of the effect of deformed core elements were prepared, analyzed and verified by hexagonal-matrix tests with 37 core elements and single row mock-up models with 7 core elements. These test results show that the rising displacements decrease with increased deformation and no rising occurs when the deformations exceed a threshold. In this paper, the effect of bending deformation due to thermal expansion and swelling on the rising displacement of the core elements was shown by seismic experiments.

Analysis of Seismic Experiment of Hexagonal Multi Bundle Model Using Core Assembly Mock-Up

2020 ◽  
Author(s):  
Akihisa Iwasaki ◽  
Shinichiro Matsubara ◽  
Hidenori Harada ◽  
Tomohiko Yamamoto
Keyword(s):  
Fast Reactor ◽  
Three Dimensional ◽  
Reactor Core ◽  
Vertical Direction ◽  
Horizontal Displacement ◽  
Vertical Displacement ◽  
The Core ◽  
Vibration Behavior ◽  
Seismic Experiment ◽  
Core Elements

Abstract The fast reactor core is composed of hundreds of core elements that stand independently on the core support plate, but does not have support to constrain vertical displacement in order to avoid effects such as thermal elongation. When the earthquake occurs, the group vibration behavior is shown, including the rising of core elements in vertical direction, the collision with adjacent core elements in horizontal direction, and the fluid structure interaction. The three dimensional core group vibration analysis code (REVIAN-3D) was constructed to evaluate them. In the case of fast reactor cores in Japan, the horizontal displacement of core elements at the outermost periphery is restricted by the core former (core barrel). However, since there is no core former in fast reactors other than Japan and the boundary conditions are different from those in Japan, the vibration behavior also differs. In this study, to grasp and estimate the group vibration behavior with and without a core former under the earthquake motion, seismic experiment of hexagonal multi bundle model using core assembly mock-up was conducted [1]. These test results show that the horizontal displacements are larger and impact force between pads of core assembly mock-up is smaller without the core former. In this paper, the analysis was verified by group vibration tests with and without a core former.

Conceptual Design and Evaluation of Residual Heat Removal System for Small Lead-Bismuth Fast Reactor

2021 ◽  
Author(s):  
Shijia Xu ◽  
Qinglong Wen ◽  
Shenhui Ruan ◽  
Ningning Zhao ◽  
Yukang Liu
Keyword(s):  
Mass Flow Rate ◽  
Conceptual Design ◽  
Mass Flow ◽  
Flow Rate ◽  
Fast Reactor ◽  
Heat Removal ◽  
The Core ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

Abstract A high efficient and reliable residual heat removal system (RHRS), which is of great importance in the development of Lead-Bismuth Cooled Fast Reactor (LBFR), was conceptually designed in present study. Based on the design of the RHRS and LBFR, the RELAP5 4.0 code is used to model the system, and then the numerical calculation of steady and transient state was carried out to obtain the important thermal-hydraulic characteristic parameters. Meanwhile, the variations of the parameters were obtained during the transient process, such as the fuel cladding temperature and the natural circulation mass flow rate. The results show that the mass flow rate of the core finally stabilizes at 3.9 kg/s, which is about 1.35% of the rated flow. The peak cladding temperature is less than 750.3 K within 72 h during the whole process, which is far below the temperature safety limit. Therefore, it can be considered that the RHRS can successfully remove the core decay heat of LBFR. This research lays a solid technical foundation for the conceptual design of the RHRS.

MTC 3D: A Fully Coupled Methodology for MSLB Safety Analysis

2014 ◽  
Author(s):  
Christian Royère ◽  
Michel Gonnet ◽  
Brahim Manchid
Keyword(s):  
Power Distribution ◽  
Three Dimensional ◽  
Reactor Core ◽  
Nucleate Boiling ◽  
Steam Line ◽  
Second Step ◽  
Inlet Temperature ◽  
Original Design ◽  
The Core ◽  
Strong Asymmetry

The main steam line break (MSLB) is an overcooling accident that may lead to an over criticality, and so to a power increase, after the reactor trip. The most penalizing single failure is a RCCA bank stuck out of the core when the reactor trip occurs. This configuration leads to a strong asymmetry of the radial power shape combined with a strong asymmetry of the core inlet temperature that results in a strongly distorted 3D power distribution. In the original design, the MSLB accident was studied with a simplified and conservative 0D method. The point kinetics approach requires the use of extremely conservative assumptions in order to account for the asymmetry in the core region that takes place during the transient. The use of the coupling between a three-dimensional neutronic code (SMART), a 3D core thermal-hydraulic code (FLICA cf. ref [4]) and a reactor coolant system code (MANTA cf. ref [3]) allows representing the 3D heterogeneity of the power shape and also of the resulting cross flows. In addition, this coupling allows determining moderator and Doppler feedback effects in a much more realistic way thus limiting accident consequences estimated. A methodology, called MTC3D (for Méthode Totalement Couplée 3D in French), has been developed using the coupling between the three codes to perform the MSLB analysis. The physical dominant parameters of the transient are identified through a comprehensive sensitivity analysis. Then, a deterministic approach is used in the entire transient simulation considering dominant parameters in a penalizing way. In a first step, neutronic data are determined with SMART calculations. In a second step, MANTA/SMART/FLICA transients are performed with penalized neutronic and thermal-hydraulic data. In a third step, as the steam line break transient is a relatively slow transient, the core power distribution is evaluated with a steady state SMART/FLICA calculation without penalization. In a last step, safety criteria, such as minimum DNBR (Departure from Nucleate Boiling Ratio) are calculated with FLICA calculations based on core power distribution calculated at the third step and boundary conditions calculated at the second step. The use of 3D neutronic and detailed thermal-hydraulic codes to model the reactor core allows considering a more physical representation of the core configuration for transient analysis. The coupling between 3D neutronic and core thermal-hydraulic codes allows exhibiting intrinsic margins without over penalizations related to a simplified 0D method.

scholarly journals MODELLING ASTRID-LIKE SODIUM-COOLED FAST REACTOR WITH SERPENT-DYN3D CODE SEQUENCE

2021 ◽  
Vol 247 ◽  
pp. 02028
Author(s):  
Wojciech Rydlewicz ◽  
Emil Fridman ◽  
Eugene Shwageraus
Keyword(s):  
Monte Carlo ◽  
Fuel Assembly ◽  
Fast Reactor ◽  
Power Distribution ◽  
Fast Reactors ◽  
The Core ◽  
Fuel Assemblies ◽  
Complex Core ◽  
Nodal Diffusion ◽  
Full Core

This study explores the feasibility of applying the Serpent-DYN3D sequence to the analysis of Sodium-cooled Fast Reactors (SFRs) with complex core geometries, such as the ASTRIDlike design. The core is characterised by a highly heterogeneous configuration and was likely to challenge the accuracy of the Serpent-DYN3D sequence. It includes axially heterogeneous fuel assemblies, non-uniform fuel assembly heights and large sodium plena. Consequently, the influence of generation and correction methods of various homogenised, few-group crosssections (XS) on the accuracy of the full-core nodal diffusion DYN3D calculations is presented. An attempt to compare the approximate time effort spent on models preparation against the accuracy of the result is made. Results are compared to reference full-core Serpent MC (Monte Carlo) solutions. Initially, XS data was generated in Serpent using traditional methods (2D single assemblies and 2D super-cells). Full core calculations and MC simulations offered a moderate agreement. Therefore, XS generation with 2D fuel-reflector models and 3D single assembly models was verified. Super-homogenisation (SPH) factors for XS correction were applied. In conclusion, the performed work suggests that Serpent-DYN3D sequence could be used for the analysis of highly heterogeneous SFR designs similar to the studied ASTRID-like, with an only small penalty on the accuracy of the core reactivity and radial power distribution prediction. However, the XS generation route would need to include the correction with SPH factors and generation of XS with various MC models, for different core regions. At a certain point, there are diminishing returns to using more complex XS generation methods, as the accuracy of full-core deterministic calculations improves only slightly, while the time effort required increases significantly.

Development and Basic Verification of Decay Heat Removal Analysis Code of Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Ping Song ◽  
Dalin Zhang ◽  
Tangtao Feng ◽  
Shibao Wang ◽  
Yapei Zhang ◽  
...  
Keyword(s):  
Experimental Data ◽  
Fast Reactor ◽  
Heat Removal ◽  
Code Verification ◽  
Design And Optimization ◽  
Decay Heat ◽  
Removal Capacity ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Pool Type

As one of the generation IV reactors, pool-type Sodium-cooled Fast Reactors (SFRs) is attracting more and more attention. Loss of flow and heat sink accident is one of the most serious accidents for SFRs. Therefore, the decay heat removal capacity after emergency shutdown is of great importance. This paper has developed a one-dimensional code named Decay heat Removal Analysis Code for Sodium-cooled Fast Reactor (DRAC-SFR) to analyze the decay heat removal capacity. In the code, the decay heat removal system contains the primary loop, the intermediate loop and air circuit. The decay heat is removed out step by step with the above three loops. Many studies have been conducted on code verification. The international benchmark analysis of EBRII reactor is applied in the code verification. The calculation is compared with the experimental data and the results of DRAC-SFR agreed well with the experimental data. The comparison with the steady state of China Experimental Fast Reactor (CEFR) shows a good agreement with the design value. The errors of all the compared parameters are within 2%. What’s more, calculation is performed to analyze the characteristics of the decay heat removal capacity for CEFR. Thus, code verification shows that DRAC-SFR is proper to evaluate the decay heat removal capacity for SFRs and has the ability to provide references and technical supports for the design and optimization of the pool-type sodium-cooled fast reactor.

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