A Design Method to Isothermalize the Core of High-Temperature Gas-Cooled Reactors

1987 ◽  
Vol 78 (3) ◽  
pp. 207-215
Author(s):  
Makoto Takano ◽  
Kazuhiro Sawa
Keyword(s):  
High Temperature ◽  
Design Method ◽  
The Core ◽  
High Temperature Gas

scholarly journals Conceptual Core Design Study of the Very High Temperature Gas-Cooled Reactor (VHTR) Upgrading the Core Performance by Using Multihole-Type Fuel

2008 ◽  
Vol 7 (1) ◽  
pp. 32-43 ◽  
Author(s):  
Kazutaka OHASHI ◽  
Tetsuo NISHIHARA ◽  
Kazuhiko KUNITOMI ◽  
Masaaki NAKANO ◽  
Yujiro TAZAWA ◽  
...  
Keyword(s):  
High Temperature ◽  
Design Study ◽  
The Core ◽  
Conceptual Core ◽  
Very High ◽  
Type Fuel ◽  
High Temperature Gas

scholarly journals Study on Air Ingress Mitigation Methods in the Very High Temperature Gas Cooled Reactor (VHTR)

2011 ◽  
Author(s):  
Chang H. Oh ◽  
Eung S. Kim
Keyword(s):  
High Temperature ◽  
Root Cause ◽  
The Core ◽  
Graphite Oxidation ◽  
Simulation Results ◽  
Oxidation Damage ◽  
Air Ingress ◽  
Mitigation Methods ◽  
Very High ◽  
High Temperature Gas

An air-ingress accident followed by a pipe break is considered as a critical event for a very high temperature gas-cooled reactor (VHTR) safety. Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break leading to oxidation of the in-core graphite structure. Thus, without mitigation features, this accident might lead to severe exothermic chemical reactions of graphite and oxygen depending on the accident scenario and the design. Under extreme circumstances, a loss of core structural integrity may occur along with excessive release of radiological inventory. Idaho National Laboratory under the auspices of the U.S. Department of Energy is performing research and development (R&D) that focuses on key phenomena important during challenging scenarios that may occur in the VHTR. Phenomena Identification and Ranking Table (PIRT) studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Oh et al. 2006, Schultz et al. 2006). Consequently, the development of advanced air ingress-related models and verification and validation (V&V) requirements are part of the experimental validation plan. This paper discusses about various air-ingress mitigation concepts applicable for the VHTRs. The study begins with identifying important factors (or phenomena) associated with the air-ingress accident using a root-cause analysis. By preventing main causes of the important events identified in the root-cause diagram, the basic air-ingress mitigation ideas can be conceptually derived. The main concepts include (1) preventing structural degradation of graphite supporters; (2) preventing local stress concentration in the supporter; (3) preventing graphite oxidation; (4) preventing air ingress; (5) preventing density gradient driven flow; (6) preventing fluid density gradient; (7) preventing fluid temperature gradient; (7) preventing high temperature. Based on the basic concepts listed above, various air-ingress mitigation methods are proposed in this study. Among them, the following one mitigation idea was extensively investigated using computational fluid dynamic codes (CFD) in terms of helium injection in the lower plenum. The main idea of the helium injection method is to replace air in the core and the lower plenum upper part by buoyancy force. This method reduces graphite oxidation damage in the severe locations of the reactor inside. To validate this method, CFD simulations are addressed here. A simple 2-D CFD model was developed based on the GT-MHR 600MWt as a reference design. The simulation results showed that the helium replaces the air flow into the core and significantly reduces the air concentration in the core and bottom reflector potentially protecting oxidation damage. According to the simulation results, even small helium flow was sufficient to remove air in the core, mitigating the air-ingress successfully.

scholarly journals Source Term Analysis of the Irradiated Graphite in the Core of HTR-10

2017 ◽  
Vol 2017 ◽  
pp. 1-6 ◽  
Author(s):  
Xuegang Liu ◽  
Xin Huang ◽  
Feng Xie ◽  
Fuming Jia ◽  
Xiaogui Feng ◽  
...  
Keyword(s):  
High Temperature ◽  
Source Term ◽  
Pressurized Water Reactor ◽  
Sample Collection ◽  
Analytical Methodology ◽  
Pressurized Water ◽  
The Core ◽  
Irradiated Graphite ◽  
Term Analysis ◽  
High Temperature Gas

The high temperature gas-cooled reactor (HTGR) has potential utilization due to its featured characteristics such as inherent safety and wide diversity of utilization. One distinct difference between HTGR and traditional pressurized water reactor (PWR) is the large inventory of graphite in the core acting as reflector, moderator, or structure materials. Some radionuclides will be generated in graphite during the period of irradiation, which play significant roles in reactor safety, environmental release, waste disposal, and so forth. Based on the actual operation of the 10 MW pebble bed high temperature gas-cooled reactor (HTR-10) in Tsinghua University, China, an experimental study on source term analysis of the irradiated graphite has been done. An irradiated graphite sphere was randomly collected from the core of HTR-10 as sample in this study. This paper focuses on the analytical procedure and the establishment of the analytical methodology, including the sample collection, graphite sample preparation, and analytical parameters. The results reveal that the Co-60, Cs-137, Eu-152, and Eu-154 are the major γ contributors, while H-3 and C-14 are the dominating β emitting nuclides in postirradiation graphite material of HTR-10. The distribution profiles of the above four nuclides are also presented.

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.

Effects of the containment chamber of a high-temperature gas-cooled reactor on the core

1985 ◽  
Vol 59 (2) ◽  
pp. 655-658
Author(s):  
A. M. Bogomolov ◽  
A. V. Zhirnov ◽  
V. A. Zavorokhin ◽  
A. S. Kaminskii ◽  
V. V. Paramonov ◽  
...  
Keyword(s):  
High Temperature ◽  
The Core ◽  
High Temperature Gas

Study on the Core Cooling Scheme After Accident Shutdown of the Pebble-Bed Modular High Temperature Gas-Cooled Reactor

2012 ◽  
Author(s):  
Yanhua Zheng ◽  
Lei Shi ◽  
Fubing Chen
Keyword(s):  
High Temperature ◽  
Temperature Distribution ◽  
Normal Operation ◽  
Helium Temperature ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Decay Heat ◽  
The Core ◽  
Core Cooling ◽  
High Temperature Gas

One of the most important properties of the modular high temperature gas-cooled reactor is that the decay heat in the core can be carried out solely by means of passive physical mechanism after shutdown due to accidents. The maximum fuel temperature is guaranteed not to exceed the design limitation, so as to the integrity of the fuel particles and the ability of retaining fission product will keep well. Nonetheless, the auxiliary active core cooling should be design to help removing the decay heat and keeping the reactor in an appropriate condition effectively and quickly in case of reactor scram due to any transient and the main helium blower or steam generator unusable. Based on the preliminary design of the 250 MW pebble-bed modular high temperature gas-cooled reactor, assuming that the core cooling will be started up 1 hour after the scram, different core cooling schemes are studied in this paper. After the reactor shutdown, a certain degree of natural convection will come into being in the core due to the non-uniform temperature distribution, which will accordingly change the core temperature distribution and in turn influence the outlet hot helium temperature. Different cooling flow rates are also analyzed, and the important parameters, such as the fuel temperature, outlet hot helium temperature and the pressure vessel temperature, are studied in detail. A feasible core cooling scheme, as well as the reasonable design parameters could be determined based on the analysis. It is suggested that, considering the temperature limitation of the structure material, the coolant flow direction should be same as that of the normal operation, and the flow rate could not be too large.

Modelling for Resuspension of Multilayer Particles

2016 ◽  
Author(s):  
Tianqi Zhang ◽  
Wei Peng ◽  
Suyuan Yu
Keyword(s):  
High Temperature ◽  
Safety Evaluation ◽  
Important Work ◽  
Modified Model ◽  
The Core ◽  
Dust Deposit ◽  
Lag Effect ◽  
Resuspension Rate ◽  
Monolayer Model ◽  
High Temperature Gas

The resuspension of graphite dust in High Temperature Gas-cooled Reactors is not only an essential way to estimate the concentration of particles in the normal operating core, but also a key phenomenon related to the release of radioactive products in depressurization accidents and the safety evaluation of environment. Precise resuspension modelling is an important work in safety analysis of HTRs. Current models were based on a single particle sitting on the substrate, however, multilayers of graphite dust deposit in the core during decades of operation. This paper uses recursion to deduce Rock’n’Roll model on each layer, and finally get the resuspension fraction changing with time and flow velocity. The results showed a “lag effect” of resuspension compared with those of Biasi’s modified model. Compared with the results of multilayer resuspension, the original monolayer model overestimated the resuspension rate and fraction.

scholarly journals Design method and research of high temperature gas floating ring seal with slightly variable gap

2021 ◽  
Vol 1081 (1) ◽  
pp. 012005
Author(s):  
Ren-jie Shi ◽  
Shuang-xi Li ◽  
Rao Zhen ◽  
Li-jun Ma ◽  
Jing-bo Zhang
Keyword(s):  
High Temperature ◽  
Design Method ◽  
Ring Seal ◽  
High Temperature Gas

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.

scholarly journals Extension of lifespan of graphite in fuel blocks of high-temperature gas-cooled reactors as the resource for ensuring design values of nuclear fuel burn-up

2019 ◽  
Vol 5 (4) ◽  
pp. 289-295 ◽  
Author(s):  
Olga I. Bulakh ◽  
Oleg K. Kostylev ◽  
Vladimir N. Nesterov ◽  
Eldar K. Cherdizov
Keyword(s):  
High Temperature ◽  
Nuclear Fuel ◽  
Alternative Energy ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Calculated Data ◽  
Fuel Burnup ◽  
The Core ◽  
High Temperature Reactors ◽  
High Temperature Gas

High-temperature gas-cooled reactor (HTGR) is one of promising candidates for new generation of nuclear power reactors. This type of nuclear reactor is characterized with the following principal features: highly efficient generation of electricity (thermal efficiency of about 50%); the use of high-temperature heat in different production processes; reactor core self-protection properties; practical exclusion of reactor core meltdown in case of accidents; the possibility of implementation of various nuclear fuel cycle options; reduced radiation and thermal effects on the environment, forecasted acceptability of financial performance with respect to cost of electricity as compared with alternative energy sources. The range of output coolant temperatures in high-temperature reactors within the limits of 750–950 °C predetermines the use of graphite as the structural material of the reactor core and helium as the inert coolant. Application of graphite ensures higher heat capacity of the reactor core and its practical non-meltability. Residence time of reactor graphite depends on the critical value of fluence of damaging neutrons (neutrons with energies above 180 keV). In its turn, the value of critical neutron fluence is determined by the irradiation temperature and flux density of accompanying gamma-radiation. The values of critical fluence for graphite decrease within high-temperature region of 800–1000 °C to 1·1022 – 2·1021 cm–2, respectively. The compactness of the core results in the increase of the fracture of damaging neutrons in the total flux. These circumstances predetermine relatively low values of lifespan of graphite structures in high-temperature reactors. Design features and operational parameters of GT-MHR high-temperature gas-cooled reactor are described in the present paper. Results of neutronics calculations allowing determining the values of damaging neutron flux, nuclear fuel burnup and expired lifespan of graphite of fuel blocks were obtained. The mismatch between positions of the maxima in the dependences of fuel burnup and exhausted lifespan of graphite in fuel blocks along the core height is demonstrated. The map and methodology for re-shuffling fuel blocks of the GT-MHR reactor core were developed as the result of analysis of the calculated data for ensuring the matching between the design value of the fuel burnup and expected total graphite lifespan.

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