scholarly journals Properties and microstructures of a matrix graphite for fuel elements of pebble-bed reactors after high temperature purification at different temperatures

Carbon ◽  
2022 ◽  
Vol 186 ◽  
pp. 739
Author(s):  
Xiang-wen Zhou ◽  
Kai-hong Zhang ◽  
Yang Yang ◽  
Lei Wang ◽  
Jie Zhang ◽  
...  
Keyword(s):  
High Temperature ◽  
Pebble Bed ◽  
Pebble Bed Reactors ◽  
Matrix Graphite ◽  
Different Temperatures ◽  
Fuel Elements

Gas-Cooled Modular Reactors With Spherical Fuel Elements — Their Inherent Safety and Applications Not Just for Power Generation

2014 ◽  
Author(s):  
Walter Jaeger ◽  
H. J. Hamel ◽  
Heinz Termuehlen
Keyword(s):  
Power Generation ◽  
High Temperature ◽  
Reactor Design ◽  
Inherent Safety ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Pebble Bed Reactors ◽  
Fuel Elements ◽  
High Temperature Gas

The gas-cooled reactor design with spherical fuel elements, referred to as high-temperature gas-cooled reactors (HTGR or HTR reactors) or pebble bed reactors has been already suggested by Farrington Daniels in the late 1940s; also referred to as Daniels’ pile reactor design. Under Rudolf Schulten the first pebble bed reactor, the 46MWth AVR Juelich reactor (Atom Versuchs-Reactor Jülich) was built in the late 1960s. It was in operation for 22 years and extensive testing confirmed its inherent safety.

The Gasification of Graphite Matrix in Pebble Bed Reactors

2009 ◽  
Author(s):  
Xinli Yu ◽  
Suyuan Yu
Keyword(s):  
High Temperature ◽  
Corrosion Rate ◽  
Fuel Element ◽  
Pebble Bed ◽  
The Core ◽  
Graphite Matrix ◽  
Matrix Graphite ◽  
Element Matrix ◽  
Fuel Elements ◽  
High Temperature Gas

This paper mainly deals with the simulations of graphite matrix of the spherical fuel elements by steam in normal operating conditions. The fuel element matrix graphite was firstly simplified to an annular part in the simulations. Then the corrosions to the matrix graphite in 10 MW High Temperature Gas-cooled Reactor (HTR-10) and the High Temperature Gas-cooled Reactor—–Pebble-bed Module (HTR-PM) were investigated respectively. The results showed that the gasification of fuel element matrix graphite was uniform and mainly occurred at the bottom of the core in both of the reactors in the mean residence time of the spherical fuel elements. This was mainly caused by the designed high temperature at the bottom. The total mass gasified in HTR-PM was much greater than the HTR-10, while it did not mean much severer corrosion occurred there. As it is known the core volume of HTR-PM is much larger than the HTR-10, which will result in much greater consumed graphite even for the same corrosion rate. The steam only lost about 1 to 3 percent after flowing through the cores in both reactors for different steam conditions. The corrosion of graphite became worse when the steam concentrations increased in helium coolant. The results also indicated that the corrosion rate of fuel element matrix graphite tended to increase slightly with the prolonging of the service time.

Fuel elements of the high temperature pebble bed reactor

1975 ◽  
Vol 34 (1) ◽  
pp. 93-108 ◽  
Author(s):  
L. Wolf ◽  
G. Ballensiefen ◽  
W. Fröhling
Keyword(s):  
High Temperature ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Fuel Elements

Wear behavior of graphitic matrix of fuel elements used in pebble-bed high-temperature gas-cooled reactors against steel

2018 ◽  
Vol 328 ◽  
pp. 353-358 ◽  
Author(s):  
Bin Wu ◽  
Yue Li ◽  
Hong-sheng Zhao ◽  
Shuang Liu ◽  
Bing Liu ◽  
...  
Keyword(s):  
High Temperature ◽  
Wear Behavior ◽  
Pebble Bed ◽  
Fuel Elements ◽  
High Temperature Gas

Deep Burn Strategy for Pure Reactor-Grade Plutonium in Pebble Bed High Temperature Gas-Cooled Reactors

2008 ◽  
Author(s):  
Eben Mulder ◽  
Dawid Serfontein ◽  
Eberhard Teuchert
Keyword(s):  
High Temperature ◽  
Fuel Cycle ◽  
Waste Incineration ◽  
Β Decay ◽  
Pebble Bed ◽  
Pebble Bed Reactors ◽  
Advanced Fuel ◽  
Very High ◽  
High Temperature Gas

In this article an advanced fuel cycle for pebble bed reactors is introduced that can safely and efficiently incinerate pure reactor-grade Pu [Pu(LWR)], thereby fulfilling the bulk of the GNEP waste incineration requirements. It is shown below that the very high fissile content of the Pu(LWR)-fuel enables it to convert practically all of the 240Pu to 241Pu and incinerate it. Since the fuel contains no 238U, no fresh 239Pu is produced. The 239Pu is reduced in-situ by 99.5% and the 240Pu by 97.6%. The only significant fissile isotope remaining is 241Pu, however, it will decay with a half life of 14.4 years to the fertile 241Am by β-decay.

Radiation Shielding Design of High Temperature Reactor Pebble-Bed Module (HTR-PM)

2016 ◽  
Author(s):  
Sida Sun ◽  
Sheng Fang ◽  
Hong Li
Keyword(s):  
High Temperature ◽  
Dose Calculation ◽  
Radiation Shielding ◽  
Source Analysis ◽  
Pebble Bed ◽  
Pebble Bed Reactors ◽  
Protection Strategies ◽  
Reduction Methods ◽  
High Temperature Reactor ◽  
Shielding Design

Radiation safety is an important concern in the design and licensing of the 200MWe High Temperature Reactor Pebble-bed Module (HTR-PM) demonstration power plant in China. To meet the requirement of the regulatory, various radiation protection strategies and methods are applied in the design process of systems and components of HTR-PM. In this study, the radiation shielding design of HTR-PM is reviewed, which includes the radiation source analysis, in-house dose calculation tool, shielding and dose reduction methods used for primary systems. The underlying conservative assumption is also discussed for correctly evaluating the dose calculation result. This summary provides a relatively systematic review of the radiation shielding methods in the design phase of HTR-PM, which may provide useful information and experiences for the radiation shielding design of future pebble-bed reactors.

Numerical Analysis of the Residual Heat Removal for the Spent Fuel Canister of HTR-PM in Fuel Loading Process

2014 ◽  
Author(s):  
Jinhua Wang ◽  
Bing Wang ◽  
Bin Wu
Keyword(s):  
High Temperature ◽  
Natural Ventilation ◽  
Maximum Temperature ◽  
Spent Fuel ◽  
Fuel Loading ◽  
Pebble Bed ◽  
Loading Process ◽  
Accident Conditions ◽  
Fuel Elements ◽  
Residual Heat

High Temperature Gas Cooled Reactor (HTGR) has inherent safety, and has been selected as one of the candidates for the Gen-IV nuclear energy system. In china, the project of the High Temperature Reactor Pebble bed Module (HTR-PM) is in design and construction process. Spherical fuel elements are chosen for the HTR-PM and the spent fuel elements will be stored in canister. The spent fuel canister will be delivered to wells for storage when fully loaded. The canister is covered by a steel cask for radiation shielding, and the cask is covered by a boron polyethylene sleeve to absorb neutrons from decay in fuel loading process. Normally, the residual heat is discharged by forced ventilation in fuel loading process. An auxiliary fan is set on top of the cask considering the possible mechanical failure for the operating fan. When losing normal power supply, the emergency power will be provided to the fans by the two line diesel generators respectively. In extreme conditions of mechanical failure for both fans, the residual heat could be discharged by natural ventilation. The temperature profiles of the different structures were studied in this paper with CFD method for both normal and accident conditions. The calculation results showed that, the maximum temperature of all of the structures are lower than the safety temperature limits in either normal or accident conditions; the temperature decreases rapidly with radial distance in the canister, and the maximum temperature is located at the center of the fuel pebble bed. So it is feasible to remove the residual heat of the spent fuel by natural ventilation in accident condition, and in the natural ventilation condition, the maximum temperature of the spent fuel, the canister shell, the shielding cask, and the boron polyethylene sleeve are lower than their safety temperature limits.

scholarly journals Computational Fluid Dynamics Applied to Study Coolant Loss Regimes in Very High Temperature Reactors

2021 ◽  
Vol 9 (2A) ◽  
Author(s):  
Uebert Gonçalves Moreira ◽  
Dany Sanchez Dominguez ◽  
Leorlen Yunier Rojas Mazaira ◽  
Carlos Alberto Brayner de Oliveira Lira ◽  
Carlos Rafael García Henández
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
High Temperature ◽  
Reactor Core ◽  
Normal Operation ◽  
Pebble Bed ◽  
Continuous Increase ◽  
High Temperature Reactors ◽  
Fuel Elements ◽  
Very High

The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this pers-pective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its   inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). We obtain the temperature profile distribution in the core for regimes where the coolant flow rate is smaller than recommended in a normal operation. In general, the temperature distributions calculated are consistent with phenomenological behavior. Even without considering the reactivity changes to reduce the reactor power or other safety mechanisms, the maximum temperatures do not exceed the recommended limits for TRISO fuel elements.

Sign in / Sign up

Export Citation Format

Share Document