scholarly journals Study on the Curved Channel Model in the Initial Core Loading of Pebble Bed High Temperature Reactors

2021 ◽  
Vol 2048 (1) ◽  
pp. 012025
Author(s):  
B Xia ◽  
J Zhang ◽  
J Guo ◽  
C Wei ◽  
Fu Li
Keyword(s):  
High Temperature ◽  
Channel Model ◽  
Reactor Core ◽  
Flow Speed ◽  
Fuel Loading ◽  
Curved Channel ◽  
Pebble Bed ◽  
Loading Process ◽  
Flow Channels ◽  
Pebble Flow

Abstract Continuous on-line fuel cycling is the essential feature of the pebble bed high temperature reactor (PB-HTR). The flow speed of the fuel pebbles in a PB-HTR presents a radial distribution in the reactor core, mainly due to the friction between the pebbles and the wall and the conical structure at the core bottom. In the VSOP fuel shuffling model, the simulation of unequal pebble flow speed is achieved by dividing the reactor core into some vertical flow channels with different numbers of the equal-volume regions in each channel. However, the fuel shuffling with equal-volume batches bring complexity when dealing with the change of fuel composition, such as the fuel fraction of fuel-graphite pebble mixture, during the initial core loading and early running-in phase. In this work, a curved channel model with unequal flow speed and the bottom cone is established based on the DEM simulation of pebble flow in the HTR-PM. The batch-wise fuel shuffling strategy is adapted to fit the complex situation during mixing and re-assigning the discharged fuels by employing a rounding strategy for the actual volume of fuels with similar irradiation history. The key of the adapted strategy is to divide the total number of the mixed batches with similar irradiation history by the number of flow channels, and round the quotient as the number of reloaded batches in each top region. The fuel loading process to build up the initial core, accompanied by the low-power reactor running to compensate the reactivity provided by the fresh fuels, is simulated by using the fuel shuffling model mentioned above. On the other hand, the simulation on the same process with an effective cylindrical core mesh and straight flow channels is carried out, in which dividing and rounding the batch numbers are unneeded. The results of both models are compared, indicating that the curved channel model presents less core reactivity and shorter fuel loading period than those of the cylindrical model. From the point of view of fidelity, the former is more suitable for the simulation of initial core loading process. The results in this work are important for enhancing the economy of fuel cycling of PB-HTRs.

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 A review of pebble flow study for pebble bed high temperature gas-cooled reactor

2019 ◽  
Vol 1 (3) ◽  
pp. 159-176 ◽  
Author(s):  
Shengyao Jiang ◽  
Jiyuan Tu ◽  
Xingtuan Yang ◽  
Nan Gui
Keyword(s):  
High Temperature ◽  
Pebble Bed ◽  
Pebble Flow ◽  
Flow Study ◽  
High Temperature Gas

Thermal Modeling of the HTR-10 Using the RELAP5-3D Code

2014 ◽  
Author(s):  
Maria Elizabeth Scari ◽  
Antonella Lombardi Costa ◽  
Claubia Pereira ◽  
Clarysson Alberto Mello da Silva ◽  
Maria Auxiliadora Fortini Veloso
Keyword(s):  
High Temperature ◽  
Reactor Core ◽  
Initial Study ◽  
Industrial Applications ◽  
Pebble Bed ◽  
State Behavior ◽  
Energy Agency ◽  
Helium Gas ◽  
New Energy ◽  
High Temperature Gas

Several efforts have been considered in the development of the modular High Temperature Gas cooled Reactor (HTGR) planned to be a safe and efficient nuclear energy source for the production of electricity and industrial applications. In this work, the RELAP5-3D thermal hydraulic code was used to simulate the steady state behavior of the 10 MW pebble bed high temperature gas cooled reactor (HTR-10), designed, constructed and operated by the Institute of Nuclear and New Energy Technology (INET), in China. The reactor core is cooled by helium gas. In the simulation, results of temperature distribution within the pebble bed, inlet and outlet coolant temperatures, coolant mass flow, and others parameters have been compared with the data available in a benchmark document published by the International Atomic Energy Agency (IAEA) in 2013. This initial study demonstrates that the RELAP5-3D model is capable to reproduce the thermal behavior of the HTR-10.

A Nodal Dynamic Model for Control System Simulation of the HTR-PM Reactor Core

2010 ◽  
Author(s):  
Zhe Dong ◽  
Xiaojin Huang ◽  
Liangju Zhang
Keyword(s):  
Control System ◽  
High Temperature ◽  
Power Distribution ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Pebble Bed ◽  
Nodal Model ◽  
High Temperature Gas

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.

scholarly journals Correction to: A review of pebble flow study for pebble bed high temperature gas-cooled reactor

2021 ◽  
Author(s):  
Shengyao Jiang ◽  
Jiyuan Tu ◽  
Xingtuan Yang ◽  
Nan Gui
Keyword(s):  
High Temperature ◽  
Open Access ◽  
Pebble Bed ◽  
Open Access Publication ◽  
Internet Portal ◽  
Creative Commons ◽  
Link Type ◽  
Pebble Flow ◽  
Flow Study ◽  
High Temperature Gas

The article “A review of pebble flow study for pebble bed high temperature gas-cooled reactor” written by Shengyao Jiang, Jiyuan Tu, Xingtuan Yang, and Nan Gui, was originally published electronically on the publisher’s internet portal (currently SpringerLink) on 11 June 2019 without open access. After publication in Volume 1, Issue 3, page 159–176, the author(s) decided to opt for Open Choice and to make the article an open access publication. Therefore, the copyright of the article has been changed to © The Author(s) 2020 and the article is forthwith distributed under the terms of the Creative Commons Attribution 4.0 International License (http://creativecommons.org/licenses/by/4.0/), which permits use, duplication, adaptation, distribution and reproduction in any medium or format, as long as you give appropriate credit to the original author(s) and the source, provide a link to the Creative Commons license, and indicate if changes were made.

Characteristics of the 250MW Pebble-Bed Modular High Temperature Gas-Cooled Reactor in Depressurized Loss of Coolant Accidents

2008 ◽  
Author(s):  
Yanhua Zhengy ◽  
Lei Shi
Keyword(s):  
High Temperature ◽  
Power Distribution ◽  
Safety Margin ◽  
Reactor Core ◽  
Reactor Power ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Loss Of Coolant ◽  
High Temperature Gas ◽  
Residual Heat

Depressurized loss of coolant accident (DLOCA) is one of the most important design basis accidents for high temperature gas-cooled reactors. Analysis of the reactor characteristic behavior during DLOCA can provide useful reference to the physics, thermo-hydraulic and structure designs of the reactor core. In this paper, according to the preliminary design of the 250MW Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), three cases of DLOCA: a instantaneous depressurization along with a flow coastdown and scram at zero time, a main pipe with a diameter of 65mm rupture, and a instrument pipe with a diameter of 10mm broken, are studied by the help of two different kinds of software THERMIX and TINTE. The key parameters of different cases including reactor power, temperature distribution of the core and pressure vessel, and the decay power removal by the passive residual heat remove system (RHRS) are compared in detail. Some uncertainties, such as residual heat calculation, power distribution, heat conductivity of fuel element, etc., are analyzed in order to evaluate the safety margin of the maximum fuel temperature during DLOCA. The calculating results show that, the decay heat in the DLOCA can be removed from the reactor core solely by means of physical processes in a passive way, so that the temperature limits of fuel and components are still obeyed. It also illustrates that the HTR-PM can reach 250MW reactor power per unit and still can keep the inherent safety.

scholarly journals Bypass flow in small absorber sphere channels of the high-temperature gas-cooled reactor pebble-bed module

2017 ◽  
Vol 10 (3) ◽  
pp. 128-139 ◽  
Author(s):  
Ziping Liu ◽  
Zeguang Li ◽  
Jun Sun
Keyword(s):  
High Temperature ◽  
Network Model ◽  
Reactor Core ◽  
Flow Distribution ◽  
Maximum Temperature ◽  
Bypass Flow ◽  
Pebble Bed ◽  
Flow Network ◽  
Control Rod ◽  
High Temperature Gas

In the high-temperature gas-cooled reactor pebble-bed module, the helium bypass flow among graphite blocks cannot be ignored due to its effect on the temperature distribution as well as the maximum temperature in the reactor core. Bypass flow was previously analyzed in the discharging tube, in vertical gaps between graphite reflectors, and in control rod channels. The focus of this study is on the bypass flow that connects the small absorber sphere channels. Different from bypass flow connecting the control rod channels, there was no evident inlet or outlet flow paths into or out of the small absorber sphere channels at the top or bottom of the reactor core. Therefore, the bypass flow connecting the pebble bed with the small absorber sphere channels was mainly caused by the horizontal gaps, in which those gaps would also be irregular due to installation, thermal expansion, or irradiation of the graphite reflectors. After clarifying the resistant coefficients of those gaps by computational fluid dynamic tools, the bypass flow distribution was calculated by the flow network model including the flow in the reactor core, small absorber sphere channels, as well as horizontal gaps. Cases with various size combinations of gaps were adopted into the flow network model to test the sensitivity of bypass flow distribution to those parameters. Finally, the bypass flow in the small absorber sphere channels was concluded to be not significant in the reactor core.

scholarly journals Double Heterogeneity Neutronic Simulation of HTR-10 for First Criticality and Full Power Initial Core with TORT-TD code

2021 ◽  
Vol 927 (1) ◽  
pp. 012037
Author(s):  
Daddy Setyawan
Keyword(s):  
High Temperature ◽  
Cross Section ◽  
Power Distribution ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Pebble Bed ◽  
The Core ◽  
Full Power ◽  
Double Heterogeneity ◽  
High Temperature Gas

Abstract In order to support the verification and validation of computational methods and codes for the safety assessment of pebble bed High-Temperature Gas-cooled Reactors (HTGRs), the calculation of first criticality and full power initial core of the high-temperature pebble bed reactor 10 MWt (HTR-10) has been defined as one of the problems specified for both code-to-code and code-to-experiment benchmarking with a focus on neutronics. HTR-10 Experimental facility serves as the source of information for the currently designed high-temperature gas-cooled nuclear reactor. It is also desired to verify the existing codes against the data obtained in the facility. In HTR-10, the core is filled with thousands of graphite and fuel pebbles. Fuel pebbles in the reactor consist of TRISO particles, which are embedded in the graphite matrix stochastically. The reactor core is also stochastically filled with pebbles. These two stochastic geometries comprise the so-called double heterogeneity of this type of reactor. In this paper, the first criticality and the power distribution in full power initial core calculations of HTR-10 are used to demonstrate treatment of this double heterogeneity using TORT-TD and Serpent for cross-section generation. HTR-10 has unique characteristics in terms of the randomness in geometry, as in all pebble bed reactors. In this technique, the core structure is modeled by TORT-TD, and Serpent is used to provide the cross-section in a double heterogeneity approach. Results obtained by TORT-TD calculations are compared with available data. It is observed that TORT-TD calculation yield sufficiently accurate results in terms of initial criticality and power distribution in full power initial core of the HTR-10 reactor.

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.

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