Allegro: The European Gas Fast Reactor Demonstrator Project

2009 ◽  
Author(s):  
Christian Poette ◽  
Vale´rie Brun-Magaud ◽  
Franck Morin ◽  
Jean-Franc¸ois Pignatel ◽  
Richard Stainsby ◽  
...  
Keyword(s):  
High Temperature ◽  
Fast Reactor ◽  
Heat Removal ◽  
Reactor System ◽  
Large Temperature ◽  
The Core ◽  
Intermediate Heat Exchanger ◽  
Current Design ◽  
Ceramic Fuel ◽  
High Temperature Components

In the Gas Fast Reactor development plan, ALLEGRO is the first necessary step towards the electricity generating prototype GFR. The ALLEGRO start of operation is planned by 2020. This needs to define all design options in 2010 and to start detailed design studies in 2013. ALLEGRO is a low power Gas Cooled Fast Reactor studied in the European framework. It is a loop type, non electricity generating reactor. Its power is about 80 MW. Several objectives are assigned to ALLEGRO. At first, it will demonstrate the viability of the GFR reactor system, no reactor of this type having been built in the past. Most of the GFR architecture, materials and components features are considered at reduced scale in ALLEGRO, excluding the energy conversion system. ALLEGRO will rely on the same safety options as the reactor system. In addition, the ALLEGRO core will allow the progressive qualification of the GFR ceramic fuel, with the possibility to load some ceramic carbide or nitride sub-assemblies in a first MOX core, with SiC/SiCf cladding and wrappers. When such unit test will be considered convincing enough, the diagrid and circuits are designed to accept full high temperature ceramic cores. The core neutrons can also be used to irradiate structural materials with fast neutron spectrum and in a large temperature range. The core can also include innovative irradiation fuel devices (samples or full bundles) for other reactor systems. Finally, branches on the main intermediate heat exchanger will allow the testing and validation of high temperature components and processes. The pre-conceptual design of ALLEGRO is shared between European partners through the GCFR 6th R&D Framework Program. After recalling the role of the European partners in the different design and safety tasks, the paper will give an overview of the current design with recent progresses in various areas like: • Core design and neutron performances, • The design of experimental advanced ceramic GFR fuel sub-assemblies included in several locations of the MOX core, • Fuel handling principles and solutions, • System design and global reactor architecture which is largely influenced by the Decay Heat Removal strategy (DHR) for depressurized accidents.

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.

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 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.

CFD Calculations of Natural Circulation in a High Temperature Gas Reactor Following Pressurized Circulator Shutdown

2011 ◽  
Author(s):  
Yu-Hsin Tung ◽  
Richard W. Johnson
Keyword(s):  
High Temperature ◽  
Free Convection ◽  
Fluid Dynamic ◽  
Natural Circulation ◽  
Flow Paths ◽  
The Core ◽  
Intermediate Heat Exchanger ◽  
Multiple Paths ◽  
Core Depth ◽  
High Temperature Gas

It is anticipated that in the event of the failure of the gas circulator in a prismatic gas-cooled very high temperature gas reactor (VHTR), there will develop natural convection currents in the core with the helium coolant. It is of interest to know the amount of energy transported by the helium plumes impinging on material surfaces in the upper plenum. Additionally, in the event of a rupture in an intermediate heat exchanger which contains water, it will be of great interest to understand the potential for free convection as it will convect water vapor, which will have detrimental effects on the core graphite. It is well known that heating a gas causes it to rise because the buoyant forces overcome gravitational forces. In the reactor, there will be hot walls that can provide heating to the helium, though the temperature of the coolant channel walls will be a function of the core depth, which makes the presence of free convection dependent on the particular conditions. In addition to the uncertainty of whether there will be sufficient buoyant forces to drive free convection, there is uncertainty as to what paths the helium will take in forming natural circulation loops. Computational fluid dynamic (CFD) calculations are reported herein that demonstrate the potential for the occurrence of natural circulation considering the core itself along with upper and lower plena and including flow paths in the gaps between the graphite blocks that allow bypass flow to occur. It is shown that multiple paths are possible for circulating flow.

Validation of THACS for Sodium Cooled Fast Reactor Based on Benchmark Analysis of EBR-II

2014 ◽  
Author(s):  
Yue Nina ◽  
Ma Zaiyong ◽  
Hu Benxue ◽  
Suizheng Qiu ◽  
Guanghui Su
Keyword(s):  
Fast Reactor ◽  
Channel Model ◽  
Flow Characteristics ◽  
Primary Loop ◽  
Reactor Kinetics ◽  
Multiple Channel ◽  
Benchmark Analysis ◽  
The Core ◽  
Intermediate Heat Exchanger ◽  
Transient Thermal

In this paper the thermal-hydraulic characteristics of the primary loop of the Experimental Breeder Reactor (EBR-II), including the temperature and the flow characteristics of the core, the intermediate heat exchanger (IHX) and the experiment subassembly XX09 and XX10, were analyzed with the transient thermal-hydraulic code THACS. The THACS code contains the core, the pumps, IHX, the sodium pool and some other modules, and each module could operate separately. All of the primary–loop components are simulated one-dimensional, and in the core calculation the incompressible model for the single phase. The multiple-channel model is applied to simulate the core subassemblies, including the average, hot, XX09, XX10, the reflector and the blanket channels. The neutron physics is calculated with the point reactor kinetics, and the reactivity feedbacks caused by the Doppler effect, coolant density, axial expansion of fuel rods and radial expansion of core are considered. Two tests, namely the SHRT-17 and SHRT-45R tests, are simulated to validate our tools and models. The THACS simulation results show that the EBR-II type sodium cooled fast reactor could shut down automatically relying on inherent negative feedbacks in the two tests.

Construction of a High Temperature Grade 91 Sodium Component Test Loop

2010 ◽  
Vol 654-656 ◽  
pp. 416-419
Author(s):  
Hyeong Yeon Lee ◽  
Jae Han Lee ◽  
Tae Ho Lee ◽  
Jae Hyuk Eoh Lee ◽  
Tae Joon Kim ◽  
...  
Keyword(s):  
Heat Exchanger ◽  
Fast Reactor ◽  
Large Scale ◽  
Performance Test ◽  
Heat Removal ◽  
Test Facility ◽  
Decay Heat ◽  
Test Loop ◽  
Intermediate Heat Exchanger ◽  
Grade 91

A large scale sodium test facility of ‘CPTL’(Component Performance Test Loop) for simulating thermal hydraulic behavior of the Korean demonstration fast reactor components such as IHX(Intermediate Heat Exchanger), DHX(Decay Heat Removal Heat Exchanger) and sodium pump under development by KAERI is to be constructed. The design temperature of this test loop is 600°C and design pressure is 1MPa. The three heat exchangers are made of Grade 91 steel. Another sodium test facility of the ‘STEF’(Sodium Thermal-Hydraulic Experimental Facility) will be constructed next to the CPTL facility to simulate the passive decay heat removal behavior in the sodium cooled fast reactor. In this paper, the overall facility features of the CPTL and STEF are introduced and preliminary conceptual design of the facilities are carried out.

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.

The Measurement of the High Temperature Sodium Pipe in CEFR

2013 ◽  
Author(s):  
Huajin Yu ◽  
Long Tang ◽  
Min Qi ◽  
Yuanwu Ye
Keyword(s):  
High Temperature ◽  
Fast Reactor ◽  
Nuclear Reactor ◽  
Heat Removal ◽  
Measure Data ◽  
Secondary Circuit ◽  
Thermal Displacement ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Displacement Measurements

The main causes of invalidation of heat transfer piping are stress, thermal displacement and vibration in fast reactor. Strain and displacement measurements are of particular importance in the commissioning and running periods, which is to ensure safety of nuclear reactor. The defect and abnormity of sodium piping can be discovered betimes by the strain and displacement measurements, and the hidden problems of piping can be avoided. Based on the investigation of measurement on piping of power plant, the strain and thermal displacement measurements are carried into execution on the main cooling systems of secondary circuit and the accident residual heat removal system of China Experimental Fast Reactor(CEFR) during the commissioning and running periods. According to the measure data, the accurateness of pipe stress analysis is validated, and the security of systems can be safeguarded. This paper describes the strain and displacement measurement on the high-temperature sodium piping, including the measure method for the double-layer piping, the test of detection capability of high temperature strain gauges, displacement measure device, the analysis and adjustment method of measure data, and so on.

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