Bed High Temperature Gas-cooled Reactor Research Articles

Numerical study of the particle-scale heat transfer in the HTR-PM pebble bed based on GPU-DEM

An essential aspect of the design of a pebble bed high-temperature gas-cooled reactor (HTGR) is the consideration of heat transfer. A particle-scale heat transfer model was developed based on the previous GPU-DEM numerical model. This model includes the heat conduction of the particles in contact, the radiative heat transfer among particles, and unidirectional convection heat transfer in the flow field. The comparison of the predicted results with the results of the TF-PBEC experiment demonstrates the accuracy of the conduction and radiation modules within the current model. Based on the current model and the porous media model, a numerical study of the natural convection heat transfer in the High-Temperature gas-cooled Reactor-Pebble-bed Module (HTR-PM) after reactor shutdown is performed. This paper systematically introduces the model design, pebble bed generation, flow field calculation, and heat transfer simulation. The results indicate that natural convection contributes to removing residual heat in the reactor core, and a higher pressure leads to better convection heat transfer. At extremely low Reynolds numbers (Re < 10), convection heat transfer is comparable to conduction and radiation, while for pressurized pebble beds (Re = 50–500), convection heat transfer dominates. Furthermore, the influence of different models used to calculate the Nusselt number of the results is compared and analyzed. For most cases, the differences in the temperature distribution of the pebble bed obtained from the different correlations are minor. However, the impact of porosity on the local Nusselt number cannot be ignored in the extremely low Reynolds numbers flows. Overall, the particle-scale heat transfer model can provide more effective reference information for the design and construction of an HTGR.

An inversion model for burnup history of spent fuels in pebble bed HTGR based on optimization algorithm

The measurement of transuranic isotopes in spent nuclear fuel is a crucial technique for nuclear material accounting and nuclear non-proliferation. In a pebble bed high-temperature gas-cooled reactor, numerous fuel spheres circulate through the reactor core multiple times, making it unfeasible to document the power histories of individual fuel spheres. This complexity renders it impossible to employ traditional methods, such as solving the forward burnup equations used in conventional reactors, to calculate the transuranic isotopes.In this approach, optimization algorithms were coupled with the data obtained from the BUrnup Measurement System to establish a power history inversion model, which was used to solve for the power history of fuel spheres and subsequently calculate the transuranic isotopes.The paper also compared the pros and cons of coupling the inversion model with different optimization algorithms and evaluated models using various schemes of activity vector. This process helped determine the most suitable optimization algorithm and scheme of activity vector. Sensitivity analysis revealed that the established inversion model exhibited a degree of tolerance to noise in input data and the underlying physical model.

Investigating geometry adjustments for enhanced performance in a PeLUIt-10 MWt pebble bed HTGR with OTTO refueling scheme

This research investigates the impact of changes in reactor geometry to discover more optimal dimensional options for nuclear power plants with potential applications in remote areas of Indonesia. The reactor used is PeLUIt-10, a pebble bed High-Temperature Gas-cooled Reactor (HTGR) with a 10 MWt capacity and a “once-through-then-out” or OTTO fuel loading scheme. The neutronic and thermal–hydraulic performances of PeLUIt-10 were analyzed with its core size altered, first by modifying its height-to-diameter ratio (H/D ratio) and second by reducing its core volume. These analyses were performed using the Pebble Bed Reactor Neutron Diffusion (PEBBED) Code, a specialized tool to analyze HTGR reactor physics parameters, particularly Pebble Bed Reactors.. Neutronic parameters analyzed include total fuel flow, burnup, power peaking factor, and power density distribution. For thermal-hydraulics and safety, parameters include steady-state and transient fuel temperatures, especially in Depressurized Loss of Forced Cooling (DLOFC) accidents. Results show that the optimal design, maintaining a volume of 5 m3, is a reactor with a height of 159.57 cm and a diameter of 200 cm (H/D ratio of 0.8). The fuel can achieve a maximum burnup of 80.97 MWD/kg-HM at this size. Power density distribution at these dimension is better than another dimensions, with a relatively low power peaking value of 1.4. For safety parameters, the fuel temperature in transient conditions remains below the safety limit. Meanwhile, when the core volume is reduced, the minimum burnup target of 60 MWD/kg-HM at a reactor volume of 4.4 m3 with a diameter of 172 cm and height of 189.2 cm. Thus, for OTTO-cycle PeLUIt-10, altering H/D ratio is more beneficial.

Thermal-Hydraulic Analysis of Pebble Bed High Temperature Gas-Cooled Reactor After Shutdown Based on FLUENT Software

In order to provide a reliable tool for thermal-hydraulic simulation of pebble bed high temperature gas-cooled reactors (HTGRs), a two-dimensional model was developed based on the porous media model and user-defined scalar (UDS) function of FLUENT software. Then, the model was applied to the numerical simulation of the shutdown test of the 10 MW high temperature gas-cooled test reactor (HTR-10) at 9 MW power level, and the temperature distribution and flow field distribution in the reactor were obtained and compared with the results of the experimental data. The reliability of the model in this paper was verified. Based on the model, the effects of the water-cooled panel temperature and the initial core temperature on the thermal-hydraulic characteristics of HTR-10 after shutdown were further explored. The results show that there is a decoupling phenomenon between the residual heat transfer within the core and the heat dissipation of the pressure vessel. The initial core temperature has relatively little effect on the heat dissipation and maximum temperature of the pressure vessel, but it has a significant impact on the thermal characteristics of the core area.

Study on on-line temperature measurement technology for core of pebble bed high temperature gas-cooled reactor

The core of the pebble bed high temperature gas-cooled reactor HTR is filled with a large number of spherical elements, which flow downward from the top of the core while the reactor is in operation. Because of the mixing ratio of spherical elements and the uncertainty of pebble flow, some (hot spots) possibly appear as a result of fuel pebble aggregation in local parts of the core. At present, the analyzation of this issue is not sufficient precise in theory, therefore on-line temperature measurement is of great significance to the improvement of HTR core thermal analysis model. In this paper, a technical scheme for online measurement of HTR core temperature is presented. The temperature monitors, with built-in metal fuse wires and the same shape as normal graphite pebble, are employed to participate in the circulation of pebbles in core. The core temperature is measured and recorded when the monitors flow through the hot spots. Based on the proposed core temperature measurement technique, the temperature and distribution in HTR-10 core were actually measured effectively and the technique was verified by practical application.

A morphological image processing method for simultaneous scrutinization of particle position and velocity in pebble flow

The flow field characteristic is vital to the design work and safety requirements of the pebble bed high temperature gas-cooled reactor (HTGR). It is necessary to study and analyze the particle location and velocity profile in pebble flow. A digital image processing (DIP) method is proposed to investigate the characteristics of particle motion, which contains two sections: mathematical morphology and threshold segmentation. These two theories are used to extract the particle position and binarize the particle image captured in the experiment, respectively. The post-processing of particle image can accurately obtain the center coordinates and velocity distribution of particles with the given time interval. Generally, the proposed algorithm can maintain a recognition accuracy of particles in pebble flow more than 90%. The calculation of particle velocity has shown that flow velocity in the middle region is much faster than in the near-wall region, and approximately forms a Gaussian distribution. The specific image processing and its conclusion can be used for dealing with relevant images and providing some reference for reactor design.

Random geometry capability in RMC code for explicit analysis of polytype particle/pebble and applications to HTR-10 benchmark

With the increasing demands of high fidelity neutronics analysis and the development of computer technology, Monte Carlo method is becoming more and more attractive in accurate simulation of pebble bed High Temperature gas-cooled Reactor (HTR), owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. For the double-heterogeneous geometry of pebble bed, traditional Monte Carlo codes can treat it by explicit geometry description. However, packing methods such as Random Sequential Addition (RSA) can only produce a sphere packing up to 38% volume packing fraction, while Discrete Element Method (DEM) is troublesome and also time consuming. Moreover, traditional Monte Carlo codes are difficult and inconvenient to simulate the mixed and polytype particles or pebbles. A new random geometry method was developed in Monte Carlo code RMC to simulate the particle transport in polytype particle/pebble in double heterogeneous geometry systems. This method was verified by some test cases, and applied to the full core calculations of HTR-10 benchmark. The reactivity, temperature coefficient and control rod worth of HTR-10 were compared for full core and initial core in helium and air atmosphere respectively, and the results agree well with the benchmark results and experimental results. This work would provide an efficient tool for the innovative design of pebble bed, prism HTRs and molten salt reactors with polytype particles or pebbles using Monte Carlo method.

DEM study of granular discharge rate through a vertical pipe with a bend outlet in small absorber sphere system

Absorber sphere pneumatic conveying is a special application of pneumatic conveying technique in the pebble bed High Temperature Gas-Cooled Reactor (HTGR or HTR). Granular discharge through a vertical pipe with a bend outlet is one of the control modes to determine solid mass flowrate which is an important parameter for the design of absorber sphere pneumatic conveying. Granular discharge rate through the vertical pipe with a bend outlet in the small absorber sphere system are investigated by discrete element method simulation. The effect of geometry parameters on discharge rate, the discharge rate fluctuation in the vertical pipe, and the effect of friction on steady discharge rate (Ws) are analyzed and discussed. The phenomena of discharge rate fluctuation in the vertical pipe are observed, which are mainly resulted from the evolution of the average downward granular velocity. The steady discharge rate decreases rapidly with sliding friction coefficient increasing from 0.125 to 0.5, and gradually saturates with the friction coefficient further increasing from 0.5 to 1. It is interesting that the linear relation between Ws2/5 and pipe internal diameter D with zero intercept are found for the vertical pipe discharge with a bend outlet, which is different from the orifice discharge through a hopper or silo with none-zero intercept. A correlation similar to Beverloo’s correlation is developed to predict the steady discharge rate through the vertical pipe with a bend outlet. These results are helpful for the design of sphere discharge rate through the vertical pipe with a bend outlet in the small absorber sphere system of the pebble bed HTGR.

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

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.

Burnup performance of OTTO cycle pebble bed reactors with ROX fuel

A pebble bed high-temperature gas-cooled reactor (PBR) with rock-like oxide (ROX) fuel was designed to achieve high discharged burnup and improve the integrity of the spent fuel in geological disposal. The MCPBR code with a JENDL-4.0 library, which developed the analysis of the Once-Through-Then-Out (OTTO) cycle in PBR, was used to perform the criticality and burnup analysis. Burnup calculations for eight cases were carried out for both ROX fuel and a UO2 fuel reactor with different heavy-metal loading conditions. The effective multiplication factor of all cases approximately equalled unity in the equilibrium condition. The ROX fuel reactor showed lower FIFA than the UO2 fuel reactor at the same heavy-metal loading, about 5–15%. However, the power peaking factor and maximum power per fuel ball in the ROX fuel core were lower than that of UO2 fuel core. This effect makes it possible to compensate for the lower-FIFA disadvantage in a ROX fuel core. All reactor designs had a negative temperature coefficient that is needed for the passive safety features of a pebble bed reactor.

Effect of bed configuration on pebble flow uniformity and stagnation in the pebble bed reactor

Pebble flow uniformity and stagnation characteristics are very important for the design of pebble bed high temperature gas-cooled reactor. Pebble flows inside some specifically designed contraction configurations of pebble bed are studied by discrete element method. The results show the characteristics of stagnation rates, recycling rates, radial distribution of pebble velocity and residence time. It is demonstrated clearly that the bed with a brachistochrone-shaped configuration achieves optimum levels of flow uniformity and recycling rate concentration, and almost no pebbles are stagnated in the bed. Moreover, the optimum choice among the arc-shaped bed configurations is demonstrated too. Detailed information shows the quantified characteristics of bed configuration effects on flow uniformity. In addition, a good design of the pebble bed configuration is suggested.

Modeling the Flow and Heat Transfer in a Packed Bed High Temperature Gas-Cooled Reactor in the Context of a Systems CFD Approach

Engineers are faced with two major challenges when carrying out the thermal-fluid design of a complex system consisting of many interacting components. The first challenge is to predict the performance of all the individual thermal-fluid components. The second challenge is to predict the performance of the integrated plant consisting of all its subsystems. The complexity associated with the thermal-fluid design of complex systems requires the use of a variety of analysis techniques and simulation tools. These range from simple one-dimensional models that do not capture all the significant physical phenomena, to large-scale three-dimensional computational fluid dynamics (CFD) codes that, for practical reasons, cannot simulate the entire plant as a single integrated model. In the systems CFD approach, a network code serves as the framework to link the models of the various components together and to control the solution. The models of the components can be of varying degrees of complexity. These can range from simple lumped models to complex fully three-dimensional CFD models. This paper gives a brief overview of the systems CFD (SCFD) approach and an overview of the model of the pebble bed nuclear reactor that was developed in the context of the SCFD approach.

ICONE19-43738 Research on Physical Characteristics of the First Core in the Pebble Bed High Temperature Gas-Cooled Reactor

ICONE19-43738 Research on Physical Characteristics of the First Core in the Pebble Bed High Temperature Gas-Cooled Reactor

06/02117 A systems CFD model of a packed bed high temperature gas-cooled nuclear reactor: du Toit, C. G. et al. International Journal of Thermal Sciences, 2006, 45, (1), 70–85.

06/02117 A systems CFD model of a packed bed high temperature gas-cooled nuclear reactor: du Toit, C. G. et al. International Journal of Thermal Sciences, 2006, 45, (1), 70–85.

Validation of a transient thermal-fluid systems CFD model for a packed bed high temperature gas-cooled nuclear reactor

This paper provides an overview of the theoretical basis for a new thermal-fluid systems CFD simulation model for high temperature gas-cooled reactors, contained in the Flownex software code. Flownex provides for detailed steady-state and transient thermal-fluid simulations of the complete power plant, fully integrated with core neutronics and controller algorithms. The reactor model is founded on a fundamental approach for the conservation of mass, momentum and energy for the compressible fluid flowing through a fixed bed, as well as the heat transfer in the pebbles and core structures. The time-wise integration of the resulting differential equations is based on an implicit pressure correction algorithm. This allows for the use of rather large time steps making it very suitable for simulating the slow transients that can be expected to follow incidents like reactor shutdowns. The paper also compares the Flownex results for four transient tests with the measured results from the SANA test facility as well as to the results of simulations with the Thermix/DIREKT code that were done at the Research Centre, Jülich. The Flownex results compare well with the Thermix/DIREKT results for all the cases presented here. Good comparison was also obtained between the simulated and measured results, except at two points within the pebble bed near the inner wall. The fact that quick computer simulation times were obtained indicates that the new model indeed achieves a fine balance between accuracy and simplicity. However, the discrepancies obtained at the two points near the inner wall, together with the fact that additional uncertainty was introduced in the original SANA test set-up by not being able to control the temperature of the outer wall, highlight the need for additional systematic tests to be performed in order to better validate the new model.

Feasibility of Burning First- and Second-Generation Plutonium in Pebble Bed High-Temperature Reactors

The core physics investigations at the Nuclear Research Consultancy Group in the Netherlands, as part of the activities within the HTR-N project of the European Fifth Framework Program, are focused on the incineration of pure (first- and second-generation) Pu fuels in the reference pebble bed high-temperature gas-cooled reactor (HTR) HTR-MODUL with a continuous reload [MEDUL, (MEhrfach DUrchLauf, multipass)] fueling strategy in which the spherical fuel elements, or pebbles, pass through the core a number of times before being permanently discharged. For pebbles fueled with different loadings of plutonium, the feasibility of a sustained fuel cycle under nominal reactor conditions was investigated by means of the reactivity and temperature coefficients of the reactor. The HTR-MODUL was found to be a very effective reactor to reduce the stockpile of first-generation plutonium. It reduces the amount of plutonium to about one-sixth of the original and reduces the risk of proliferation by denaturing the plutonium vector. For second-generation plutonium the incineration is less favorable, as the amount of plutonium is only halved.

A systems CFD model of a packed bed high temperature gas-cooled nuclear reactor

The theoretical basis and conceptual formulation of a comprehensive reactor model to simulate the thermal-fluid phenomena of the PBMR reactor core and core structures is given. Through a rigorous analysis the fundamental equations are recast in a form that is suitable for incorporation in a systems CFD code. The formulation of the equations results in a collection of one-dimensional elements (models) that can be used to construct a comprehensive multi-dimensional network model of the reactor. The elements account for the pressure drop through the reactor; the convective heat transport by the gas; the convection heat transfer between the gas and the solids; the radiative, contact and convection heat transfer between the pebbles and the heat conduction in the pebbles. Results from the numerical model are compared with that of experiments conducted on the SANA facility covering a range of temperatures as well as two different fluids and different heating configurations. The good comparison obtained between the simulated and measured results show that the systems CFD approach sufficiently accounts for all of the important phenomena encountered in the quasi-steady natural convection driven flows that will prevail after critical events in a reactor. The fact that the computer simulation time for all of the simulations was less than three seconds on a standard notebook computer also indicates that the new model indeed achieves a fine balance between accuracy and simplicity. The new model can therefore be used with confidence and still allow quick integrated plant simulations.