High-temperature Gas-cooled Reactor Core Research Articles

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.

Burnable poison optimized on a long-life, annular HTGR core

The present work presents analysis results of the core design optimizations for an annular, prismatic High Temperature Gas-cooled Reactor (HTGR) with passive decay-heat removal features. Its thermal power is 100 MWt and the operating temperature is 850 °C (1123 K). The neutronic calculations are done for the core with heterogeneous distribution of fuel and burnable poison particles (BPPs) to flatten the reactivity swing and power peaking factor (PPF) during the reactor operation as well as for control rod (CR) insertion into the core to restrain a small excess reactivity less than 1$. The next step of the study is done for evaluation of core reactivity coefficient of temperature.

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.

Computed tomography neutron detector system to observe power distribution in a core with long neutron flight path

A power distribution monitoring system by using a moving detector for a core with a long neutron flight path has been proposed. High Temperature Gas-cooled Reactor (HTGR) and Fast Reactor (FR) has a long neutron flight path and the neutrons reach to detector far from fuel assembly in the center of the core unlike Light Water Reactor (LWR). By using the feature, power distribution can be observed with a few detectors by moving the detector and computed tomography technology similar to X-ray Computed Tomography (CT). For a small-sized core, the power distribution can be evaluated only by an ex-core neutron detector. For a large-sized core with inner detectors, the power distribution can be observed with a small number of in-core detectors even if the deployment is limited due to material integrity conditions such as temperature environment. The feasibility is numerically confirmed by simulations of the HTGR core and its detector response. It is expected to observe the power distribution in the core of HTGR and FR, which is difficult continuously to deploy in-core detectors because of high temperature and/or high irradiation damage.

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.

Flattening of the Power Distribution in the HTGR Core with Structured Control Rods

Control rods (CRs) have a significant influence on reactor performance. Withdrawal of a control rod leaves a region of the core significantly changed due to lack of absorber, leading to increased fission rate and later to Xe135 buildup. In this paper, an innovative concept of structured control rods made of tungsten is studied. It is demonstrated that the radial division of control rods made of tungsten can effectively compensate for the reactivity loss during the irradiation cycle of high-temperature gas-cooled reactors (HTGRs) with a prismatic core while flattening the core power distribution. Implementation of the radial division of control rods enables an operator to reduce this effect in terms of axial power because the absorber is not completely removed from a reactor region, but its amount is reduced. The results obtained from the characteristic evolution of the reactor core for CRs with a structured design in the burnup calculation using the refined timestep scheme show a very stable core evolution with a reasonably low deviation of the power density and Xe135 concentration from the average values. It is very important that all the distributions improve with burnup.

Comparative Analysis of Genetic Algorithms Based on Different Selection Strategies for Refueling Optimization in the Ratio Method

The genetic algorithm is one of the classic algorithms applied to the refueling optimization. An important part of this algorithm is the selection strategies. The existing studies often directly adopt the roulette wheel selection and stochastic tournament selection, and are lacking in comparison and analysis of different selection strategies. To obtain the selection strategy with the strongest optimization capability, this study, with the 1/6 core of a thorium-based prismatic high-temperature gas-cooled reactor (HTGR) taken as an example, constructs the fitness function in the ratio method, performs core physics calculation using the DRAGON code, and in conjunction with the elitism strategy, compares the optimization capabilities of the five selection strategies, including the roulette wheel selection, stochastic tournament selection, uniform ranking method, exponential ranking selection and deterministic selection. The study results show that the optimization capability of the exponential ranking selection is superior to the other four strategies, so the exponential ranking selection is most suitable for solving the refueling optimization problems.

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.

DEVELOPMENT OF PIN-LEVEL NEUTRONICS/THERMAL-FLUID ANALYSIS COUPLED CODE SYSTEM FOR A BLOCK-TYPE HTGR CORE

Recently, the coupling between computer codes that simulate different physical phenomena has attracted for more accurate analysis. In the case of high-temperature gas-cooled reactor (HTGR), the coupling between neutronics and thermal-fluid analysis is necessary because of large change of temperature in the reactor core. Korea Atomic Energy Research Institute (KAERI) has developed the coupled code system between a reactor physics analysis code CAPP and a thermal-fluid system safety analysis code GAMMA+ for a block-type HTGR. The CAPP/GAMMA+ coupled code system provides more accurate block-wise distribution data than CAPP or GAMMA+ stand-alone analysis. However, the block-wise distribution data has the limitation in order to predict safety parameters such as the maximum temperature of the nuclear fuel. It is necessary to calculate refined distribution, for example, pin-level (fuel compact level) distribution. In this study, we tried to solve this problem by coupling CAPP and a high-fidelity thermal-fluid analysis code CORONA. CORONA can perform a high-fidelity thermal-fluid analysis of Computational Fluid Dynamics (CFD) level by dividing a block-type HTGR core into small lattices. On the other hand, CAPP can provide a pin power distribution. It is expected that the refined, more accurate distribution data for a block-type HTGR can be obtained by coupling these two codes. This paper presents the development of coupled code system between CAPP and CORONA, and then it is tested on a simple HTGR column problem with encouraging results.

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.

Feasibility study on burnable poison credit concept to HTGR fuel fabrication from core specification perspective

Feasibility study on Burnable Poison (BP) credit concept to High Temperature Gas-cooled Reactor (HTGR) fuel fabrication has been performed. By mixing BP into fuel material in the first place of fuel fabrication, criticality safety is ensured in the all fuel fabrication process even with high enrichment fuel such as 14 wt% used in commercial HTGR. However, the poison effect also degrades the criticality even in the HTGR core, and it may shorten cycle length and achievable burn-up of the core. Therefore, the effect is evaluated by whole core burn-up calculation. As a BP, boron, gadolinium, erbium, and hafnium are investigated. As a result, it is found that boron and gadolinium suit this concept, and the 14 wt% fuel can be fabricated even with severe Light Water Reactor (LWR) class safety measure corresponding to 5 wt%.

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.

Three-dimensional modeling and loss-of-coolant accident analysis of high temperature gas cooled reactor

The high temperature gas-cooled reactor (HTR) is well known due to its prominent inherent safety, high efficiency and low carbon environmental protection. Accurate and reliable thermal hydraulic analysis is of great significance to the system design of HTR. Based on the coolant flow and heat transfer process in the core, this study establishes a three-dimensional coupled model for the core of HTR, and the temperature variation in the core is simulated. The model is verified by comparing the calculated results with those published data by THERMIX software. After that, the models are employed to study thermal hydraulic characteristics under the loss of coolant accident. The results show that the maximum relative error of the calculated results with published data are 2.1% under full power, the three-dimensional model can be employed to simulate the flow and heat transfer process in the primary circuit of HTR. The calculation of 50% of full power shows that the model fully considers the influence of convective heat transfer and radiative heat transfer between the core of the pebble bed and the core side wall and the cold air chamber, making numerical calculations closer to actual condition. It is found when approximately 20% of the coolant is lost, the outlet temperature of helium reaches 1140 K, which exceeds the helium outlet temperature limit (1126 K). In addition, the inherent safety of an HTR be verified under transient conditions of natural convection.

Tritium Transport and Distribution in a High-Temperature Gas-Cooled Reactor

The High-Temperature Gas-cooled Reactor (HTGR) has been selected as one of the next-generation nuclear power plants because of its passive safety features. The Korea Atomic Energy Research Institute (KAERI) has been studying how to utilize HTGR efficiently and safely. The HTGR uses graphite as a moderator and helium as a coolant. Once tritium is produced, it is released into the coolant; once released from the core, tritium travels within the primary loop. Because the coolant is gas phase, it is easy to transport to other systems. While it circulates in the primary loop, tritium is involved in processes that include leakage, purification, and permeation. KAERI has been developing a tritium behavior analysis code named TRitium Overall Phenomena analYsis (TROPY) to analyze tritium transport and predict the amount of tritium in the loop in the HTGR core. In this paper, the functions of the TROPY code are introduced, and the amount of tritium in each loop and the amount released into the product hydrogen from the MHTGR 350-MW(thermal) core are explained.

Analysis of Clumped-Pebble Shape on Thermal Radiation and Conduction in Nuclear Beds by Subcell Radiation Model

Abstract The core of high-temperature gas-cooled reactor is a dense pebble bed of random packing filled with monosized fuel spheres. Subcell radiation model (SCM) is a generic analytical approach to calculate effective thermal conductivity (ETC) of thermal radiation. For the packed bed of monosized spheres operated in various conditions, it is proven that the SCM is still applicable in the particle size ranges of 1.2–60 mm and temperature ranges of 0–1200 °C. Based on the SCM, radiation-to-conduction ratio ξ is presented and radiation becomes an essential part at ξ>0.1 for the accurate evaluation. For the beds of nonoverlapping clumped-sphere particles, the model combining with discrete element method (DEM) and SCM is presented to study the heat transfer behaviors, including effects of particle shape, emissivity distribution and pebble flow with transient heat transfer. For the experimental nuclear pebble beds, the results of SCM are in good agreement with the empirical correlation and accord well with the experimental data under high temperature range.

Benchmark calculations for verification of thermo-fluid analysis codes using simplified model of the OECD/NEA MHTGR-350 core design

Sponsored by the OECD/NEA, a benchmark study on prismatic high temperature gas-cooled reactor (HTGR) using the MHTGR-350 design was initiated. One of the important lessons learned during the process of Phase 1 Exercise 2 of the OECD/NEA MHTGR-350 benchmark is that even the simplest benchmark case of Phase 1 Exercise 2 is too complex to identify reasons for differences among the results obtained by different computer codes. In this work, therefore, simpler benchmark test cases have been developed to improve the efficiency of the verification process of thermo-fluid analysis codes to simulate a prismatic HTGR core. Retaining the major physical thermo-fluid phenomena, the original OECD/NEA benchmark model of Phase 1 Exercise 2 was simplified and four test cases were established using well-organized boundary conditions. Numerical simulations using three computer codes (i.e., GAMMA+, AGREE, and CFX) were performed and good agreement was achieved among the results obtained by the three codes. It is believed that the proposed benchmarks are particularly helpful for code developers and/or researcher who attempt to solve the official MHTGR-350 Benchmark for the first time.

Neutronics Analysis for a High-Temperature Gas-Cooled Reactor in a Water-Flooded-Core Accident

A water-flooded-core accident is a serious potential accident for high-temperature gas-cooled reactors (HTGRs). In this technical note, based on two different water-flooded-core scenarios, preliminary neutronics analysis was performed on a typical HTGR. Preliminary temperature-effect analysis is carried out as well. It is found that the neutron-slowing ability is the key for the effective multiplication factor of the HTGR core. More importantly, when the water-flooded-core accident occurs, the HTGR might return back to supercritical with the core temperature decreasing even if it is safely shut down at high operation temperature.

Fully ceramic microencapsulated fuel in prismatic high-temperature gas-cooled reactors: Design basis accidents and fuel cycle cost

Fully ceramic microencapsulated (FCM) fuel may enhance the already strong inherent reactor safety characteristics of a conventional High-Temperature Gas-Cooled Reactor (HTGR). FCM fuel uses an SiC matrix that exhibits higher stability under irradiation with limited swelling relative to a conventional graphite matrix. Additionally, the SiC matrix is expected to exhibit improved mechanical performance relative to graphite. The feasibility of FCM-fueled HTGRs has been explored in a previous study (Lu et al., 2018), and three FCM fuel concepts were identified to be able to maintain the cycle length of the reference conventional General Atomics 350 MWt prismatic modular HTGR.In this work, the previous investigation of the reactor core safety characteristics of HTGRs fueled by these three FCM fuels was significantly extended to detailed thermal hydraulics and neutronics analysis of normal operation and design basis accident scenarios. We found that these FCM-fueled cores would have a maximum fuel temperature ∼20–40 K higher than that of the reference HTGR core at the beginning of cycle (BOC) and ∼30–60 K higher at the end of cycle (EOC) during normal operating conditions. Pressurized (P-) and depressurized (D-) loss of forced cooling (LOFC) accidents as well as the control rod withdrawal accident were explored. The FCM-fueled cores would at most have a maximum fuel temperature ∼30–50 K higher than that of the reference core in a P-LOFC accident, ∼10–30 K in a D-LOFC accident and ∼30–50 K in a rod withdrawal accident. The conclusion of the study is that FCM fuel has a small impact on the reactor performance and safety characteristics during normal operation and design basis accident conditions within the existing and generic 350 MWt prismatic modular HTGR design we considered.Additionally, a fuel cycle cost comparison was also performed in the current study. We found that the FCM fuels may be up to 40%–74% more expensive than the conventional fuel. Further analysis of the economics of FCM fuel are needed, because the uncertainties in the assumptions for this study are high.

High Temperature Oxidation Behavior of Nickel and Iron Based Superalloys in Helium Containing Trace Impurities

A high-temperature gas-cooled reactor (HTGR) is recognized as the best candidate reactor for next generation nuclear reactors. Helium is used to be the coolant in the core of the HTGR with temperature expected to exceed 900 °C at the core outlet. Several iron- and nickel-based superalloys, including Alloy 800H, Hastelloy X, and Alloy 617, are potential structural materials for intermediate heat exchanger (IHX) in an HTGR. Oxidation behaviors of three selected alloys (Hastelloy X, Alloy 800H, and Alloy 617) were investigated at four different temperatures from 650℃ to 950 ℃ under helium environments with various concentrations of O2 and H2O. Preliminary results showed that chromium oxide as the primary protective layer was observed on surfaces of the three tested alloys. Based on results of mass gain and SEM analyses, Hastelloy X alloy exhibited the best corrosion resistance in all corrosion tests. Further details on the oxidation mechanism of these alloys are presented in this study.

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.