Pebble Bed High Temperature 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.

Thermohydraulic Analysis of Coolant Entry Patterns in a High-Temperature Pebble Bed Reactor: Insights and Characterization

Thermohydraulic Analysis of Coolant Entry Patterns in a High-Temperature Pebble Bed Reactor: Insights and Characterization

Perturbation algorithm for refueling reactivity coefficient in the pebble-bed high temperature reactor

The fuel management mode of the pebble-bed high temperature reactor (PB-HTR) is on-line continuous refueling, that is, the reactor is refueling at the same time of operation, and the reactivity is continuously supplemented by continuous refueling. In this paper, the perturbation algorithm for refueling reactivity coefficient is proposed, and the internal mechanism of reactivity introduced by refueling in the PB-HTR is revealed, that is, refueling reactivity depends on the dot product of the pebble flow velocity and the differential of atomic density in space. Based on the neutron diffusion equation, the difference and differential formulas of refueling reactivity coefficient are derived. The application of differential formula, which can be directly calculated according to the current parameters of the reactor core before the refueling, is simpler than the difference formula. Then, the equilibrium core and running-in phase of HTR-PM are used to verify the effect of the perturbation algorithm and the calculation results agree well with the reference value from the VSOP code. The distribution of refueling reactivity in different spatial regions can also be analyzed based on the difference and differential formulas.

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.

Safeguards and Security for High-Burnup TRISO Pebble Bed Spent Fuel and Reactors

Several high-temperature thermal neutron–spectrum pebble bed reactors are being commercialized. China has started up two helium-cooled pebble bed high-temperature reactors. In the United States, the X-Energy helium-cooled and the Kairos Power salt-cooled pebble bed high-temperature reactors will produce spent nuclear fuel (SNF) with burnups exceeding 150 000 MWd per tonne. The reactor fuel in each case consists of small spherical graphite pebbles (4 to 6 cm in diameter) containing thousands of small TRISO (microspheric tri-structural isotropic) fuel particles embedded in the fuel of zone these pebbles. The unique isotopic, chemical, and physical characteristics of this high-burnup SNF create a technical case to eliminate safeguards based on the low risk for use in nuclear weapons, while maintaining safeguards in terms of risk for use in radiological weapons. These safeguards could be reduced to the simple counting and monitoring of pebbles in storage. Alternatively, there is the option to create a special category with reduced requirements for this SNF in storage, transport, and disposal. No safeguards would be required for a repository with only this type of SNF. Reactor safeguards are required for fresh fuel, partly burnt fuel, and to identify unconventional pebbles with depleted uranium or other materials that might be used to create weapons-useable materials.

Estimation of transuranic isotopes in irradiated fuel elements for pebble bed HTR on activity vector

Due to the unknown power history of fuel elements in a pebble-bed High-Temperature Reactor, the traditional method of calculating transuranic isotopes through forward burnup calculations using initial loading data is not applicable.This paper presents a novel method for estimating transuranic isotopes within fuel elements in pebble bed HTR. It involves the power history inversion model using activity vector detectable by the online Burnup Measurement System of High Temperature Gas-cooled Reactor Pebble bed Module to predict the power history of the fuel elements. Subsequently, transuranic isotopes are calculated through a forward burnup process.To check the performance of the burnup equation inverse model. Nuclear Inventory Tool were used to generate database that provides known activity vector as target and corresponding power history of a fuel element. Results show that the inverse model leads to a more stable convergence, effectively reconstructing the complex power history. The integrated power estimation error remains below 0.2%, with prediction errors for ten transuranic isotopes of interest below 1.5%.

Optimum power determination and comparison of multi-pass and OTTO fuel management schemes of HTGR-based PeLUIt reactor

A co-generation reactor based on pebble-bed High-Temperature Gas-Cooled Reactor (HTGR) has been designed. This study focuses on designing a modular nuclear reactor based on pebble bed high-temperature reactor (Electric Power and Industrial Hot Steam Plant/Pembangkit Listrik dan Uap Panas Industri, abbreviated as PeLUIt), an upgrade from Indonesia’s 10 MW Experimental Power Reactor (Reaktor Daya Eksperimen, RDE). PeLUIt is proposed as a solution for electricity and co-generation application. The 5-pass and Once-Through-Then-Out (OTTO) refueling scheme for the PeLUIt has been analyzed. PEBBED was used for equilibrium core calculations, whereas COMBINE7 was used for spectrum calculations utilizing ENDF/B-VII.0 as nuclear data. Based on the results, it can be inferred that the PeLUIt co-generation reactor analysis with an optimal power of 40 MWt is produced by increasing fuel intake into the core while retaining core geometry and fuel design. In this study, burn-up levels for each refueling scheme are also analyzed, and the value for OTTO is lower than 5-pass refueling schemes. In addition, the Depressurized Loss of Forced Cooling (DLOFC) accident scenario for both types of refueling schemes was also analyzed. The result showed that the OTTO refueling scheme has a higher safety level than the 5-pass refueling scheme.

Simulation of random medium of fuel pebble with liquid salt coolant using PebMC code

Pebble-bed High Temperature reactors (HTR) are proposed to generate heat for power production and process heat application such as Hydrogen (H2) production. A HTR fuel pebble consists of random distribution of several thousands of fuel particles, called Tri-Structural Isotropic (TRISO), in fuel zone followed by a graphite layer. This configuration along with coolant layer results in double heterogeneity effect. A multi-group Monte Carlo code PebMC has been developed to accurately analyse molten salt cooled HTR lattice cell. Neutron transport in such a complex fuel distribution requires resonance self-shielded cross-section which is generated using rational approximation and Dancoff correction. Dancoff factor accounts for those neutrons leaving a fuel particle and directly colliding with another fuel particle, without any collision with moderator nuclei.In this study, an analytical expression based on a two-region model developed for gas cooled HTRs is adopted to calculate Dancoff factor for pebble with salt coolant. The implementation of these analytical expressions with 172groups WIMS library in PebMC code have been discussed. Subsequently, paper describes neutron tracking in random medium of HTR lattice using Monte Carlo method. The influence of fuel salt on Dancoff factor for various packing of TRISO particles and subsequent effect on infinite multiplication factor (K-inf) calculation have been presented. In addition, effect of temperature of molten salt on Dancoff factor have been studied.

Design of a supervisory control system for autonomous operation of advanced reactors

Advanced reactors to be deployed in the coming decades will face deregulated energy markets, and may adopt flexible operation to boost profitability. To aid in the transition from baseload to flexible operation paradigm, autonomous operation is sought. This work focuses on the control aspect of autonomous operation. Specifically, a hierarchical control system is designed to support constraint enforcement during routine operational transients. Within the system, data-driven modeling, physics-based state observation, and classical control algorithms are integrated to provide an adaptable and robust solution. A 320MW Fluoride-cooled High-temperature Pebble-bed Reactor is the design basis for demonstrating the proposed control system.The hierarchical control system consists of a supervisory layer and low-level layer. The supervisory layer receives requests to change the system’s operating conditions (e.g., the current reactor power to meet a load-follow), and accepts or rejects them based on constraints that have been assigned. Constraints are issued to keep the plant within an optimal operating region. The low-level layer interfaces with the actuators of the system to fulfill requested changes, while maintaining tracking and regulation duties. To accept requests at the supervisory layer, the Reference Governor algorithm was adopted. To model the dynamics of the reactor, a system identification algorithm, Dynamic Mode Decomposition, was utilized. To estimate the evolution of process variables that cannot be directly measured (e.g., the propagation of delayed neutron precursors), the Unscented Kalman Filter, incorporating a nonlinear model of nuclear dynamics, was adopted. The composition of these algorithms led to a numerical demonstration of constraint enforcement during a 40% power drop transient (at a rate of 5%/min). Uncontrolled secondary-side temperatures were successfully constrained. Adaptability of the proposed system was demonstrated by modifying the constraint values, and enforcing them during the transient. Robustness was also demonstrated by enforcing constraints under noisy environments.

Measurement of Helium and Xenon Knudsen permeability in graphite IG-11

Understanding the flow and diffusion (transport) of gases through nuclear graphite is of interest in both the prismatic and pebble bed high-temperature gas reactors (HTGRs), for normal operation and under accident conditions, for example, as related to graphite oxidation. Both the laminar and turbulent gaseous transport are of interest. It has also been noted that nuclear graphite can have pore sizes that can possibly lead to non-continuum transport, as the pore size can be comparable to the gaseous mean free path. We report measurements and analyzes of experiments performed with IG-11 graphite in the pressure range of 10 Pa to 14,700 Pa, 293 K, using Helium, and research grade xenon (99.999%), respectively. These values are in fair agreement with those reported for different graphites and gases by other investigators, but are also higher by six orders of magnitude than those reported for Kr permeability in a developmental very low permeability graphite-HXT-90. Our measured ratio of the Helium and Xenon permeabilities is about 4, different from mXemHe≈5.72 that corresponds to diffuse reflection at the graphite surface for both gases. Simple analysis indicates that the two gases have slightly different accommodation coefficients (about 25%) with the graphite surface. However, there could also be other reasons, such as experimental uncertainties or the need for a more detailed analysis.

Analysis of gamma spectra from burnup measurement in HTR-10

HTR-10, the 10 MW pebble-bed High Temperature Gas-cooled Reactor-Test Module designed by Tsinghua University, achieved its first criticality in 2000. One of the unique features of HTR-10 is the continuous fuel feeding achieved by its fuel cycle system, in which the fuel sphere is allowed to pass through the core several times before discharged out. During the fuel circulation, the on-line burnup measurement system is used to ensure fuel credit. Up to now, after nearly 21 years of operation, huge data including gamma spectra have been collected by this on-line burnup measurement system. In this paper, the gamma spectra data collected from Nov. 2005 to Mar. 2006 are analyzed to identify the radionuclides left in the irradiated fuel sphere after a long decay. The analysis includes energy identification, decay scheme confirmation, activity estimation and so on. This work shows that radionuclides and corresponding activities are preliminarily identified, which lays a base for R&D of new burnup measurement apparatus and the profiling of fuel irradiation history. That would help to increase accuracy for the nuclear material accounting and to improve the operation strategy of pebble bed high-temperature gas-cooled reactors.

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

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.

Preliminary Analysis on Burnup Calculation of Several Arrangement of TRISO and Pebble inside an MCNP Model of HTR Core

Several studies related to simplifying the modeling of pebble bed High-Temperature Reactor core (HTR) has been developed before. From some calculation on several MCNP models with a fueled pebble to dummy ratio 57:43, using a combination of several types of TRISO (TRi-structural ISOtropic particle fuel) unit and Pebble unit is modeled to achieve its first criticality. In this paper, some MCNP model that uses 27000 pebbles with a 57:43 ratio and 100% fueled pebble is created to be used on burnup calculation and to compare its k-eff and nuclide inventory. From this burnup calculation, it could be seen that SC (Simple Cubic) TRISO unit has faster calculation time followed by the HCP (Hexagonal Close Packed) TRISO unit and then the FCC (Face-Centered Cubic) TRISO unit. The BCC (Body-Centered Cubic) pebble unit had some consistent deviation from another pebble unit, and it still needs more study to know more about the reason behind it. It could be seen that if there are some dummy pebbles inside the reactor, then the deviation would be higher than if there is just fueled pebble inside the reactor. On the 57:43 ratio, the absolute average deviation of k-eff on burnup calculation is lower than 2% and 10% for nuclide inventory (mass). On 100% fueled pebble, it’s below 0.15% on k-eff absolute deviation and below 8% on nuclide inventory deviation.

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

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.

Optimization of PeLUIt-150 Conceptual Design

A 150MWt pebble bed high temperature reactor, called PeLUIt-150, is proposed as a solution for distributed energy demand including the electricity and various co-generation potentials. In current study an optimization of the design is performed. The optimization includes its optimum power density to have a compact core design including the height-to-diameter (H/d) ratio. As several computer codes is used in this study, PEBBED-THERMIX codes, comparison of results from different code is also discussed. Constrain of the design at current stage are the maximum fuel temperature under depressurized loss of forced cooling (DLOFC) accident which represent the passive performance of the core, and discharge burnup which represent the fuel utilization performance of the core. Current study achieves an optimized design with core diameter and height of 260 cm and 601.11cm, a total active core volume of 31.9 m3. The average power density is 4.7W/c m3 with discharge burnup of 80.82MWd/Kg and the maximum fuel temperature after DLOFC is 1619.9°C.

An explicit expression of empirical parameter in ZBS model for predicting pebble bed effective thermal conductivity

The pebble bed effective thermal conductivity is a key parameter in the design and safety analysis of the high temperature pebble bed reactor. The ZBS model is widely used to predict the pebble bed effective thermal conductivity. However, the predictive ability of the ZBS model is greatly affected by the selected values of the empirical parameter contact area fraction φ. In this paper, a semi-empirical formula is proposed to estimate the empirical parameter contact area fraction φ, which is related with the contact radius ratio γ and coordination number Nc. The stagnant effective thermal conductivities in the three types of ordered pebble beds are numerically studied to provide the values of φ with corresponding contact radius ratios γ and coordination numbers Nc. Then the multiple linear regression is performed and the proposed semi-empirical formula is determined asφ=0.15Nc1.1γ. The predicted values by the improved ZBS model with the estimation of contact area fraction φ agree rather well with the numerical results of the random pebble bed. Compared with SANA experimental data, the improved ZBS model is proved to be better than other models. Compared with HTTU experimental data, the improved ZBS model is able to predict the effective thermal conductivity in the bulk region of pebble beds. However, the improved ZBS model needs to be further improved in the near-wall region.

CFD validation with optimized mesh using benchmarking data of pebble-bed high-temperature reactor

In this study, a computational fluid dynamics (CFD) analysis conducted to predict the thermo-hydraulic performance of the HTR-10, which is an experimental pebble bed reactor (PBR). The PBR with a modular model with a realistic approach and finite volume method (FVM) instead of porous media is used to predict the distribution of the critical thermal-hydraulics parameters like temperature, pressure, velocity, and heat transfer coefficient (HTC) over the pebbles as well as the reactor core. In the simulation study, an optimized mesh used as a replacement of the large size mesh. The simulation results, for the boundary conditions such as coolant inlet and outlet temperatures, mass flow, and other thermal-hydraulics parameters, were validated with the data published by the IAEA benchmark document. This study demonstrates that the CFD analysis using an optimized mesh is capable of predicting the nature thermo-hydraulic behavior of PBR core, maintaining design safety limits like maximum fuel temperature, and coolant flows distribution.

Pebble Bed High Temperature Reactor for Electric and Cogeneration Application

Specific energy demand of Indonesia with its distributed energy demand required a special energy supply solution. The energy supply solution should be able to fulfil the electric demand of huge and growing population which spread in many different islands also to fulfil the energy demand to process abundances yet distributed natural sources. This paper proposes a pebble bed high temperature reactor as an energy supply in the form electric and/or cogeneration application. A parametric survey of the core geometry of a 150MWt Pebble Bed Reactor called PeLUIt is performed, follow by an equilibrium analysis of the optimized design. An accident analysis of depressurized loss of forced cooling (DLOFC) is performed to show the strong passive safety features of the design. Finally the heat-mass balance of the plant is also presented.

Numerical study of natural convection effects on effective thermal conductivity in a pebble bed

The effective thermal conductivity (ETC) is a key parameter of the heat transfer models in high temperature pebble-bed reactors. The ETC in a simple cubic packed bed of graphite pebbles dominant by the passive heat transfer coupled of the conduction, radiation and natural convection is studied numerically, with the emphasis on the effects of the natural convection. It is found that the natural convection has little influence on the solid pebble temperatures but it would affect the fluid temperatures. The ETCs in the bulk region are consistently greater than ETCs in the near-wall region. The natural convection plays a positive role in ETCs in the bulk region while it has a negative role in ETCs in the near-wall region. When the temperature increases, the natural convection is less important to the ETC. The natural convection effect on the ETC could be ignored in the high temperature conditions (T > 600 K).