Gas-cooled Reactor Core Research Articles

Modular gas-cooled reactor core optimal design and additive manufacturing performance based on SiC matrix dispersed with TRi-structural ISOtropic fuel

High-temperature gas-cooled reactors present significant technical competitiveness in various fields, such as low-carbon electricity, thermal energy, and hydrogen supply. Silicon Carbide (SiC) matrix dispersed with TRi-structural ISOtropic (TRISO) particle fuel is characterized by high-temperature tolerance, radiation resistance, and good neutron economy, which has emerged as a pivotal technological pathway for reactor core design. With the rapid advancement of SiC Additive Manufacturing (AM) technology, there is a substantial increase in core design flexibility, and an innovative gas-cooled reactor design was proposed by Monte Carlo (MC) method in this paper. The Non-uniform Rational B-spline (NURBS), Latin Hypercube Sampling (LHS), second-order polynomial response surface, as well as Genetic Algorithm (GA) were coupled to realize the optimization design for the coolant channel. Besides, a modular fuel design was proposed, and the AM SiC shell was fabricated using Binder Jetting (BJ) technology. Furthermore, the densification of the AM SiC shell and the matrix with SiC powder and SiC microspheres (TRISO simulator) was investigated by Chemical Vapor Infiltration (CVI). Finally, the microscopic morphology analysis and thermal-mechanical performance characterization of the samples were conducted, which demonstrated the technical feasibility and advanced capabilities of the SiC matrix dispersed with TRISO fuel based on AM technology.

Seismic response prediction using intensity measures: Graphite nuclear reactor core model case study

Seismic response analyses of structures have conventionally used the peak ground acceleration or spectral acceleration as an intensity measure to estimate the engineering demand parameters. An extensive shaking table test program was carried out on a quarter-sized advanced gas-cooled reactor (AGR) core model to investigate the global dynamic behavior of the system with degraded graphite components while subjected to seismic excitation. Evaluation of the most widely considered intensity measures, with respect to their capability for predicting the seismic response of an AGR core–like structure, is performed. Twenty intensity measures of 16 distinct seismic input motions are formulated and correlated, with experimental measurements describing the dynamic response of the reactor core model. Linear correlations are constructed for each intensity measure to statistically determine the best metric for predicting the seismic response of the AGR core model, and statistical analysis indicates that the acceleration spectrum intensity (ASI) is best suited to characterize and describe the structural demand of an AGR core-like structure when subjected to seismic loading. A response prediction tool is developed, based on empirically derived linear correlations, to estimate column distortions and determine the critical input motion for further experimental and numerical studies. Statistical analysis indicates that predicted column distortions, compared against direct experimental displacements, are significant, repeatable, and accurate.

Numerical study on the flow and heat transfer characteristics in a high-temperature gas-cooled reactor core

In the high-temperature gas-cooled reactor (HTGR), the core operation heavily relies on the heat transfer process facilitated by the coolant that withdraws the heat released by fuel particles. Therefore, it is imperative to carefully study the flow and heat transfer characteristics between the fuel particle and coolant. Additionally, under high core temperature conditions, the fuel particle tends to incur damage. In this study, we have designed a new type of HTGR using hydrogen as a coolant, and performed numerical simulations to explore the flow and heat transfer conditions in the core under various working conditions. Our results indicate that under high velocity inlet and low temperature inlet conditions, the system provides a high heat transfer rate. Furthermore, interesting relationships were observed between the flow and heat transfer characteristics of the model through the Reynolds number (Re), n, and Tr. Since the density of the coolant changes as the temperature varies, the hydrogen is susceptible to compression. Notably, poor thermal conductivity on the fuel particle surface generates hotspots at the points of flow stagnation, flow separation, and flow convergence. Our study on the new reactor core's flow and heat transfer characteristics provides an important reference for improving the core outlet temperature and the fuel particle's structure.

A mathematical framework for evaluation of column shape profiles for an advanced gas-cooled reactor core model subject to seismic excitation

The safe shutdown of Advanced Gas-cooled Reactor (AGR) nuclear power stations in response to a seismic event is vital to their safety case. The tubular graphite bricks used to moderate the neutrons within an AGR core are arranged in columns whose bores provide channels for either fuel or control rods. Earthquake-induced distortion of the channels could impede the insertion of the control rods and compromise the safe shut-down, maintenance and servicing of the reactor. This paper presents a mathematical framework, utilising Euler mechanics, to evaluate the column shape displacement profiles of fuel and control rod channels within a state-of-the-art quarter-sized physical model of an AGR core when subjected to seismic loading. The data obtained from sensors installed within the model bricks, and configured to monitor interface displacements, are used to infer the global behaviour of a multi-stacked brick column subject to seismic excitation. Directly measured displacements of the top of the brick columns, obtained using a motion capture vision system, are compared with the displacements calculated using the framework presented, verifying the validity of the procedure. Statistical analysis is employed to quantify and characterise the performance of the Euler mechanics method. For multiple model build configurations, which represent different brick-cracking scenarios in an aged core, the Pearson correlation factor between the direct and indirect measurements is evaluated for the top of the column displacements giving an average value of 0.96 in the direction of the input motion. This shows that good agreement is achieved for the column shape displacement time-histories. The seismic responses are shown to be significantly larger in amplitude in the presence of large numbers of cracked bricks.

Development of a Low-Cost 6 DOF Brick Tracking System for Use in Advanced Gas-Cooled Reactor Model Tests.

This paper presents the design of a low-cost, compact instrumentation system to enable six degree of freedom motion tracking of acetal bricks within an experimental model of a cracked Advanced Gas-Cooled Reactor (AGR) core. The system comprises optical and inertial sensors and capitalises on the advantages offered by data fusion techniques. The optical system tracks LED indicators, allowing a brick to be accurately located even in cluttered images. The LED positions are identified using a geometrical correspondence algorithm, which was optimised to be computationally efficient for shallow movements, and complex camera distortions are corrected using a versatile Incident Ray-Tracking calibration. Then, a Perspective-Ray-based Scaled Orthographic projection with Iteration (PRSOI) algorithm is applied to each LED position to determine the six degree of freedom pose. Results from experiments show that the system achieves a low Root Mean Squared (RMS) error of 0.2296 mm in x, 0.3943 mm in y, and 0.0703 mm in z. Although providing an accurate measurement solution, the optical tracking system has a low sample rate and requires the line of sight to be maintained throughout each test. To increase the robustness, accuracy, and sampling frequency of the system, the optical system can be augmented with an Inertial Measurement Unit (IMU). This paper presents a method to integrate the optical system and IMU data by accurately timestamping data from each set of sensors and aligning the two coordinate axes. Once miniaturised, the developed system will be used to track smaller components within the AGR models that cannot be tracked with current instrumentation, expanding reactor core modelling capabilities.

Numerical Study on the Thermal Field and Heat Transfer Characteristics of a Hexagonal-Close-Packed Pebble Bed

Fuel elements in a high-temperature gas-cooled reactor (HTGR) core may be stacked with a hexagonal close-packed (HCP) structure; therefore, analyzing the temperature distribution and heat transfer efficiency in the HCP pebble bed is of great significance to the design and safety of HTGR cores. In this study, the heat transfer characteristics of an HCP pebble bed are studied using CFD. The thermal fields and convective heat transfer coefficients under different coolant inlet velocities are obtained, and the velocity fields in the gap areas are also analyzed in different planes. It is found that the strongest heat transfer is shown near the right vertices of the top and bottom spheres, while the weakest heat transfer takes place in areas near the contact points where no fluid flows over; in addition, the correlation of the overall heat transfer coefficient with the Reynolds number is proposed as havg = 0.1545(k/L)Re0.8 (Pr = 0.712, 1.6 × 104 ≤ Re ≤ 4 × 104). It is also found that the heat transfer intensity of the HCP structure is weaker than that of the face-centered-cubic structure. These findings provide a reference for reactor designers and will contribute to the development of safer pebble-bed cores.

Numerical modelling of modular high-temperature gas-cooled reactors with thorium fuel

Abstract The volumetric homogenization method for the simplified modelling of modular high-temperature gas-cooled reactor core with thorium-uranium fuel is presented in the paper. The method significantly reduces the complexity of the 3D numerical model. Hence, the computation time associated with the time-consuming Monte Carlo modelling of neutron transport is considerably reduced. Example results comprise the time evolutions of the effective neutron multiplication factor and fissionable isotopes (233U, 235U, 239Pu, 241Pu) for a few configurations of the initial reactor core.

Design and Calibration of a Hall Effect System for Measurement of Six-Degree-of-Freedom Motion within a Stacked Column.

This paper presents the design, development and evaluation of a unique non-contact instrumentation system that can accurately measure the interface displacement between two rigid components in six degrees of freedom. The system was developed to allow measurement of the relative displacements between interfaces within a stacked column of brick-like components, with an accuracy of 0.05 mm and 0.1 degrees. The columns comprised up to 14 components, with each component being a scale model of a graphite brick within an Advanced Gas-cooled Reactor core. A set of 585 of these columns makes up the Multi Layer Array, which was designed to investigate the response of the reactor core to seismic inputs, with excitation levels up to 1 g from 0 to 100 Hz. The nature of the application required a compact and robust design capable of accurately recording fully coupled motion in all six degrees of freedom during dynamic testing. The novel design implemented 12 Hall effect sensors with a calibration procedure based on system identification techniques. The measurement uncertainty was ±0.050 mm for displacement and ±0.052 degrees for rotation, and the system can tolerate loss of data from two sensors with the uncertainly increasing to only 0.061 mm in translation and 0.088 degrees in rotation. The system has been deployed in a research programme that has enabled EDF to present seismic safety cases to the Office for Nuclear Regulation, resulting in life extension approvals for several reactors. The measurement system developed could be readily applied to other situations where the imposed level of stress at the interface causes negligible material strain, and accurate non-contact six-degree-of-freedom interface measurement is required.

Гибридная «синтез-деление» реакторная установка на ториевом топливе с источником дополнительных термоядерных нейтронов

The results of full-scale numerical experiments of a hybrid thorium-containing fuel cell facility operating in a close-to-critical state due to a controlled source of fusion neutrons are discussed in this work. The facility under study was a complex consisting of two blocks. The first block was based on the concept of a high-temperature gas-cooled thorium reactor core. The second block was an axially symmetrical extended plasma generator of additional neutrons that was placed in the near-axial zone of the facility blanket. The calculated models of the blanket and the plasma generator of D-T neutrons created within the work allowed for research of the neutronic parameters of the facility in stationary and pulse-periodic operation modes. This research will make it possible to construct a safe facility and investigate the properties of thorium fuel, which can be continuously used in the epithermal spectrum of the considered hybrid fusion-fission reactor.

Maintaining the close-to-critical state of thorium fuel core of hybrid reactor operated under control by D-T fusion neutron flux

The results of full-scale numerical experiments of a hybrid thorium-containing fuel cell facility operating in a close-to-critical state due to a controlled source of fusion neutrons are discussed in this work. The facility under study was a complex consisting of two blocks. The first block was based on the concept of a high-temperature gas-cooled thorium reactor core. The second block was an axially symmetrical extended plasma generator of additional neutrons that was placed in the near-axial zone of the facility blanket. The calculated models of the blanket and the plasma generator of D-T neutrons created within the work allowed for research of the neutronic parameters of the facility in stationary and pulse-periodic operation modes. This research will make it possible to construct a safe facility and investigate the properties of thorium fuel, which can be continuously used in the epithermal spectrum of the considered hybrid fusion–fission reactor.

A hybrid porous model for full reactor core scale CFD investigation of a prismatic HTGR

Three-dimensional (3-D) Computational Fluid Dynamics (CFD) analysis of a whole nuclear reactor core is a great challenge due to the large geometric volume and complex structures. This research presents a hybrid porous (HP) model to simplify a prismatic High Temperature Gas-cooled Reactor (HTGR) core, so 3-D CFD investigation can be performed on a full reactor core scale. In the HP model, the prototypic small coolant channels in the nuclear fuel blocks are lumped together to form multiple equivalent large coolant channels, and then the porous medium flow model is applied to each of them. Therefore, heat transfer in fuel blocks is computed by a hybrid combination of solid energy and porous flow energy equations. The similarity between the HP model and prototypic model is achieved by deriving the porous flow permeability, inertial resistance factor, and artificial thermophysical properties. Compared with the widely used whole porous (WP) flow model, the HP model preserves more realistic geometric structures, and therefore more accurate physical processes. The General Atomics’ Modular High Temperature Gas-cooled Reactor (MHTGR) design was chosen as a prototype to demonstrate the methodology. Simulations were performed using the prototypic CFD model and HP model at steady-state forced circulation, steady-state natural circulation, and transient conditions that correspond to normal operation, extended period of pressurized cool down, and short-term transients after reactor shutdown, respectively. The comparison shows good agreement between the HP model and prototypic model in the maximum fuel temperature, average solid temperature, and helium flow rate, which demonstrates the potential applicability of the HP model for a full reactor core scale simulation in the future. As a benefit, the HP model reduces the mesh quantity by a factor of 50 from a prototypic model. Correspondingly, the computation time was reduced by a factor of at least 30.

Comparative study of conceptual design of gas-cooled fast reactor core type tall versus pan cake based on MCANDLE-B burn up strategy

Comparative study of the conceptual design of Gas-cooled Fast Reactor (GFR) core type tall versus pancake based on modified CANDLE-B (MCANDLE-B) burnup strategy has been done. MCANDLE-B is a burnup strategy that utilizes natural uranium or depleted fuel as its input cycle. The conceptual design of a reactor core that compared is the tall cylinder and the pancake cylinder. In this case, the fuel used is U-10%Zr, SS-316 as a cladding material and helium as a coolant. The total volume of the two reactor cores is the same, namely 15.4 m3. SRAC 2K6 software with PIJ and CITATION modules is used to carry out simulations. The PIJ module is used for fuel cell calculations and the CITATION module is used for reactor core calculations. The results of the comparison show that the pancake core allows the reactor core to use fuel with a volume fraction of 50%: 10%: 40%, for fuel, cladding and coolant respectively. The design obtained can be operated for 10 years without refueling.

Radiological characterisation of graphite components in Advanced Gas-cooled Reactor cores

As the Advanced Gas-cooled Reactors approach the end of generation, research and characterisation are required to support the decommissioning strategy. Radiological data for AGR graphite are negligible and the radiological inventory of the AGR core and other graphite components rely on activation modelling. This is the first study of C-14 activity and its release behaviour in AGR core graphite and its associated carbonaceous deposits and provides valuable information that can support decommissioning activities. In combination with corresponding studies on Magnox core graphite, significant understanding is attained on the main C-14 precursors in the graphite and the deposits.In addition, this study reports C-14, H-3 and gamma spectrometry data on AGR graphite fuel sleeves. This is a waste stream that is currently stored in heavily engineered stores at a significant cost. The data indicate that alternative storage and disposal options with a lower environmental and financial impact are worth considering.

Temperature field in gas-cooled reactor core in transient conditions under different approaches to mass flow profiling

Positive effect of profiling the gas-cooled reactor core within the framework of the GT-MHR project was investigated in (Podgorny and Kuzevanov 2017, Kuzevanov and Podgorny 2017, 2018). The necessity arises to supplement already implemented analysis of equilibrium conditions of core operation with investigation of effects of profiling on the temperature field in transient modes of reactor core operation. The present paper is dedicated to the investigation of development of transients in gas-cooled nuclear reactor core subject to the implementation of different principles of core profiling. Investigation of transients in reactor core represents complex problem, solution of which by conducting direct measurements is beyond the resources available to the authors. Besides the above, numerical simulation based on advanced CFD software complexes (ANSYS 2016, 2016a, 2016b, Shaw 1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994) is also fairly demanding in terms of required computer resources. The algorithm for calculating temperature fields using the model where the reactor core is represented as the solid medium with gas voids was developed by the authors and the assumption was made that heat transfer due to molecular heat conductivity can be described by thermal conductivity equation written for continuous medium with thermal physics parameters equivalent to respective parameters of porous object in order to get the possibility of obtaining prompt solutions of this type of problems. Computer code for calculating temperature field in gas-cooled reactor in transient operation modes was developed based on the suggested algorithm. Proprietary computation code was verified by comparing the results of numerous calculations with results of CFD-modeling of respective transients in the object imitating the core of gas-cooled nuclear reactor. The advantage of the developed computer code is the possibility of real-time calculation of evolution of conditions in complex configurations of gas-cooled reactor cores with different channel diameters. This allows using the computer code in the calculations of transients in loops of reactor facility as a whole, in particular for developing reactor simulators. Results are provided of calculations of transients for reactor core imitating the core of gas-cooled nuclear reactor within the framework of GT-MHR project performed for different approaches to profiling coolant mass flow. Results of calculations unambiguously indicate the significant difference of temperature regimes during transients in the reactor core with and without profiling and undeniable enhancement of reliability of nuclear reactor (Design of the Reactor Core 2005, International Safeguards 2014, Safety of Nuclear Power Plants 2014) with profiling of coolant mass flow in the reactor core as a whole.

Температурное поле в активной зоне газоохлаждаемого ядерного реактора в переходных режимах при различных условиях профилирования массового расхода

Температурное поле в активной зоне газоохлаждаемого ядерного реактора в переходных режимах при различных условиях профилирования массового расхода

Gas-cooled nuclear reactor core shaping using heat exchange intensifiers

The need to shape reactor cores in terms of coolant flow distributions arises due to the requirements for temperature fields in the core elements (Safety guide No. NS-G-1.12. 2005, IAEA nuclear energy series No. NP-T-2.9. 2014, Specific safety requirements No. SSR-2/1 (Rev.1) 2014). However, any reactor core shaping inevitably leads to an increase in the core pressure drop and power consumption to ensure the primary coolant circulation. This naturally makes it necessary to select a shaping principle (condition) and install heat exchange intensifiers to meet the safety requirements at the lowest power consumption for the coolant pumping. The result of shaping a nuclear reactor core with identical cooling channels can be predicted at a quality level without detailed calculations. Therefore, it is not normally difficult to select a shaping principle in this case, and detailed calculations are required only where local heat exchange intensifiers are installed. The situation is different if a core has cooling channels of different geometries. In this case, it will be unavoidable to make a detailed calculation of the effects of shaping and heat transfer intensifiers on changes in temperature fields. The aim of this paper is to determine changes in the maximum wall temperatures in cooling channels of high-temperature gas-cooled reactors using the combined effects of shaped coolant mass flows and heat exchange intensifiers installed into the channels. Various shaping conditions have been considered. The authors present the calculated dependences and the procedure for determining the thermal coolant parameters and maximum temperatures of heat exchange surface walls in a system of parallel cooling channels. Variant calculations of the GT-MHR core (NRC project No. 716 2002, Vasyaev et al. 2001, Neylan et al. 1994) with cooling channels of different diameters were carried out. Distributions of coolant flows and temperatures in cooling channels under various shaping conditions were determined using local resistances and heat exchange intensifiers. Preferred options were identified that provide the lowest maximum wall temperature of the most heat-stressed channel at the lowest core pressure drop. The calculation procedure was verified by direct comparison of the results calculated by the proposed algorithm with the CFD simulation results (ANSYS Fluent User’s Guide 2016, ANSYS Fluent. Customization Manual 2016, ANSYS Fluent. Theory Guide 2016, Shaw1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994).

Effect of inserted sphere size on heat transfer characteristics of FCC structured pebble bed in a HTGR

Hot spots appearing in an operating high temperature gas-cooled reactor (HTGR) core have been considered as the most possible reason leading to a severe accident like fission production releasing to the environment, therefore, investigation on their positions and thus seeking ways to reduce the possibility of their appearance have attracted scientists’ attention. In our previous studies, heat transfercharacteristics of a face–centered–cubic (FCC) structured pebble–bed have been discussed,and a correlation on heat transfer coefficient with Reynolds number was presented. In this study, a method, placing a small sphere in thegap area, which is able to enhance the convective heat transfer wasproposed and the effect verifiedas well. The influence of the sphere diameter on heat transfer performances wasinvestigated in details. It is concluded through results analysis that (1) inserted sphere lowered thelocal surface temperature of adjacent pebbles by varying surrounding flow field;(2) maximum velocity of the fluid and average heat transfer coefficientincreased with sphere diameter, particularly, comparing with no small sphere case, 12.95% enhancement was achieved. Such findings may provide dataand information for reactor designers, andhelp to develop a safer HTGR pebble–bedcore.

Профилирование активной зоны газоохлаждаемого ядерного реактора с использованием интенсификаторов теплообмена

Профилирование активной зоны газоохлаждаемого ядерного реактора с использованием интенсификаторов теплообмена

Thermal Mixing Performance of Mixing Structure at Reactor Outlet of HTR-PM under Complex Conditions

In order to investigate the effects of bypass flow and power change and deviation on radial temperature difference of the helium flow entering the steam generator, model experiments and simulation calculations are proposed to evaluate the Thermal Mixing Performance (TMP) of the thermal mixing structure at Pebble-bed Module High Temperature gas-cooled Reactor (HTR-PM) core outlet. The experiments on Test Facility - Hot Gas Mixing (TF-HGM) are carried out to observe the influences of bypass flow and reactor power deviation. Simulation calculations are conducted for the mixing structure of TF-HGM and HTR-PM. According to the results of experiments and CFD simulation, the maximum radial temperature difference at the outlet of the mixing structure fulfils the requirement by steam generator, under all considered conditions. In addition, the TMP obtained by the temperature difference between main flow and bypass flow is suitable for the experiment and the simulation results while the new definition of TMP integrating specific heat, flow rate, inlet position and thermal uniformity needs further improvement.

Performance of thermal mixing structure of HTR-PM regarding bypass flow and power effect

The temperature of bypass flow not flowing through the reactor core is obviously lower than that of the main coolant helium in a High Temperature Gas-cooled Reactor (HTGR). In addition, change and deviation of reactor power will introduce higher radial temperature difference of main coolant helium. In order to investigate the effects of bypass flow and power change and deviation on radial temperature difference of the helium flow, model experiments and simulation calculations are proposed to evaluate the temperature profiles and Thermal Mixing Performance (TMP) of the thermal mixing structure at Pebble-bed Module High Temperature gas-cooled Reactor (HTR-PM) core outlet. We gave three definitions of TMP by describing the thermal nonuniformity of the inlet flow of the mixing structure in different ways. The TMP defined by the temperature difference between main flow and bypass flow is within the range of 97.5 ± 1.1% under all considered conditions which is suitable for evaluating the experiment and simulation results. The new definition of TMP integrating specific heat, flow rate and inlet position needs further improvement. In addition, the maximum radial temperature difference at the outlet of the mixing structure fulfils the requirement of steam generator, less than 30 °C (±15 °C). Furthermore, future works are discussed.