Pebble-bed Module Research Articles

Study on Conventional Island Retrofit Strategies for Converting Coal-Fired Power Plants to Nuclear Power Stations in China

The conversion of coal-fired power plants to nuclear power stations is a potential method for decarbonizing coal power and offers a pathway for low-carbon development in China’s power industry. This paper focuses on retrofitting China’s coastal coal-fired power stations and compares the potential nuclear reactor technologies for the retrofit: China’s mainstream pressurized water reactor and the commercially operated fourth-generation high-temperature gas-cooled reactor (HTGR). The analysis compares the degree of matching between the two technologies and coal-fired power stations in terms of unit capacity, thermal system parameters, unit speed, structural dimensions, and weight, which significantly impact the retrofit scheme. The results indicate that HTGR is more compatible with coal-fired power plants and is recommended as the type of nuclear reactor technology to be retrofitted. The study selected the 210 MWe High-Temperature Gas-Cooled Reactor Pebble-Bed Module (HTR-PM) as the reactor technology for retrofitting a typical 300 MW class subcritical coal-fired unit. Based on the concept of subcritical parameters upgrading, the potential analysis and strategy study of retrofit is carried out in terms of the turbine, the main heat exchange equipment, the main pumps, and the main thermal system pipelines in the conventional island. The results indicate that the conventional island of the HTR-PM nuclear power plant has significant potential for retrofitting, which can be a crucial research direction for nuclear retrofitting of coal-fired power plants.

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%.

Solving the issue of reliability data for FOAK equipment in an innovative nuclear energy system

There is a demand for the improved safety and economics of advanced nuclear energy systems. The world's first high-temperature gas-cooled reactor pebble-bed module nuclear power plant (HTR-PM Demo Project) is about to be completed in China. This power plant achieved initial full power at the end of 2022. The success of the HTR-PM Demo Project shows the importance of risk-informed initiatives throughout the project with solid support from comprehensive risk assessment. This paper focuses on a technical issue that needs to be solved when conducting risk-informed initiatives for HTR-PM, namely the issue of reliability data for first-of-a-kind (FOAK) equipment. In short, no data for the FOAK equipment are available in existing reliability databases. On the basis of the HTR-PM pilot experience, this paper proposes a process of estimating the required data for the reliability parameters of FOAK equipment. The process has five steps: (1) identification of key parts and assemblies, (2) preliminary determination of the reliability parameters and the associated reference values for the key parts and assemblies, (3) analysis of maintenance management tactics, (4) estimation of reliability parameters for the FOAK equipment using a Bayesian network, and (5) Bayesian updating with operational data. The helium circulator of HTR-PM is taken as an example to demonstrate the feasibility of the proposed process. A lack of reliability data is a common issue for innovative nuclear energy systems. The proposed process is supposed to be universal to all types of innovative design.

Rapid analysis of packed pebble beds for thermal–hydraulic characteristics via reduced order models

In recent years, the pebble bed module has garnered increased attention due to its inherent safety and versatile applications, particularly in the form of High-Temperature Gas-Cooled reactors and fusion tritium breeding blankets. Understanding the thermal–hydraulic characteristics within the pebble bed is crucial for the safety evaluation and design of such reactors. To study pore flow in the pebble bed, discrete element coupling methods and computational fluid dynamics (DEM-CFD) are commonly employed. However, simulating thousands of pebbles can be computationally intensive and time-consuming. To address this, reduced order models (ROMs) have been developed to reduce the computational complexity by employing simplified systems instead of the original complex system, thereby significantly saving computational time while maintaining accuracy. In this work, ROMs have been established for body-centered cubic packing, face-centered cubic packing, and randomly packed beds of pebbles. Using the ROM results, subsequent analysis focused on two important thermal–hydraulic characteristics: pressure and effective thermal conductivity. A comparison was made between the results obtained from the full-order model (FOM) using CFD and the reconstructed ROM results. The case studies demonstrate the potential of the proposed approach in rapidly and accurately investigating significant flow and heat transfer features, such as pressure and temperature.

Global Model Calibration of High-Temperature Gas-Cooled Reactor Pebble-Bed Module Using an Adaptive Experimental Design

The world’s first high-temperature gas-cooled reactor pebble-bed module (HTR-PM) nuclear power plant adopts an innovative reactor type and a modular structure design. Parameter estimation and model calibration are of great significance prior to the implementation of model-based control and optimization. This paper focuses on identifying the thermal hydraulic parameters of HTR-PM over the global operating domain. The process technology and model mechanism of HTR-PM are reviewed. A parameter submodel named global parameter mapping is presented to quantify the relationship between an unknown model parameter and different operating conditions in a data-driven manner. The ideal construction of such a mapping requires reliable estimates, a well-poised sample set and an appropriate global surrogate. An adaptive model calibration scheme is designed to tackle these three issues correspondingly. First, a systematic parameter estimation approach is developed to ensure reliable estimates via heuristic subset selection consisting of estimability analysis and reliability evaluation. To capture the parameter behavior among the multiple experimental conditions and meanwhile reduce the operating cost, an adaptive experimental design is employed to guide condition testing. Experimental conditions are sequentially determined by comprehensively considering the criteria of sampling density, local nonlinearity and parameter uncertainty. Support vector regression is introduced as the global surrogate due to its capability of small-sample learning. Finally, the effectiveness of the model calibration scheme and its application performance in HTR-PM are validated by the simulation results.

Numerical study on heat transfer characteris tics of supercritical water in straight and helical tubes

The subcritical steam generators in high-temperature gas-cooled reactor pebble-bed modules can be replaced by supercritical steam generators to improve thermal efficiency. The axial and circumferential nonuniform flow and heat transfer characteristics of supercritical water in straight and helical tubes with varying inclinations (0–90°) are simulated to investigate the combined effects of tube structure, flow direction, thermophysical properties, property variations, buoyancy, and centrifugal forces. With heat fluxes of 233–930 kW/m2 and inlet mass fluxes of 225 or 1260 kg/(m2 s) under a working pressure of 24.5 MPa, the friction coefficients, turbulent kinetic energy, secondary flow, temperature, and heat transfer coefficients are studied in the cross section of variable structure tubes. The wall temperature and heat transfer coefficients of four regions are compared in straight and helical tubes; the maximum heat transfer coefficient appeared before the pseudo-critical point. The heat transfer coefficients are the greatest at the bottom of the straight tubes, where fluidity was high, and at the outer of the helical tubes, where the fluid velocity was high under centrifugal force. The secondary flow in the helical tubes causes a nonuniformity in the wall temperature, causing the circumferential maximum wall temperature to move.

Heat transfer of supercritical pressure water in helical tubes

Subcritical steam generators in the current high-temperature gas-cooled reactor pebble-bed module in China can be replaced by supercritical steam generators to better coordinate the reactors and supercritical steam turbine unit to improve the thermal efficiency with no fluid phase change at supercritical pressures. The heat transfer of supercritical-pressure water flow in helical tubes with various helix diameters was simulated using the re-normalization group k–ε model to investigate the effects of finite-thickness walls, inlet mass flux, and helix diameter. The finite-thickness wall barely affected local heat transfer in the straight tube with inlet mass flux of 1,260 kg/(m2 s) whereas the local heat transfer in the helical tube was very different with relative difference in q/qav between the inside and bottom, caused by the centrifugal force, is 36%. In helical tubes, the heat transfer coefficients in the inner bar were the lowest, with a low fluid velocity under the effect of the centrifugal force. The buoyancy and centrifugal force in the helical tubes caused non-uniformity in the wall temperature; however, when the fluid temperature increased below the pseudo-critical temperature, the temperature distributions showed an increasing trend. The non-uniformities in the wall temperature distributions decreased as the centrifugal force increased, and they were suppressed by the finite-thickness wall.

Identification-based sensor and actuator fault diagnosis for industrial control systems and its application to HTR-PM

This paper studies sensor and actuator fault diagnosis for a class of linear time-invariant (LTI) industrial control systems. Based on the dynamic model obtained using system identification, a method is proposed that can diagnose all sensor and actuator faults in the control system when they do not happen simultaneously. First, a structured residual design method is developed for coupled systems. Then model errors that occur in system identification are considered and the robust isolability is analyzed. Finally, online decision rules are given which are easy to implement. The method does not require fault data and can deal with small faults that cannot be detected by inspecting input–output signals. Verification in the high temperature gas-cooled reactor pebble-bed module (HTR-PM) simulation platform shows good potential of the method for complex industrial control systems.

An event sequence modeling method in multi-unit probabilistic risk assessment for high temperature gas-cooled reactor

The number of multi-unit nuclear power plants (NPPs) increases, so does the demand for multi-unit probabilistic risk assessment (MUPRA). Considering dependencies among multiple units on one site, conventional probabilistic risk assessment (PRA) for a single unit cannot meet requirements of the overall site risk analysis. Existing MUPRA studies mainly evaluate limited risk characteristics, such as core damage, or focus on a limited number of units, within the range of two to six units. This study proposes a multi-unit event sequence modeling method for MUPRA, and develops a corresponding computer-aided tool, making it possible to deal with considerable number of units. Furthermore, their application to high temperature gas-cooled reactor – pebble bed module (HTR-PM) was carried out, showing that the modeling method and computer-aided tool can be applied to address dependencies of function events (FEs) and provide a reference for the full risk spectrum.

Fracture behavior analysis of brittle graphite sphere using a special bond model

High-Temperature Reactor–Pebble-bed Module (HTR-PM) takes the spherical fuel element design. Failure or not of fuel sphere would highly affect system operation and nuclear safety. Crush strength of fuel sphere is an important quality control measure, which is mainly determined by the graphite matrix. A bonded particle model (BPM) from discrete element method (DEM) is disclosed, validated, and applied in the fracture analysis of brittle graphite sphere. On the basis of validated BPM, the breakage behavior of nuclear fuel sphere under high-speed impact, that is, dynamic loading, is revealed in detail by a time-marching numerical simulation. This meshfree particle method can be effectively extended to the failure computation of different composite structures (e.g., particle-reinforced metal matrix composite) by defining heterogeneous bonds inside distinct material interface.

600-MWe high-temperature gas-cooled reactor nuclear power plant HTR-PM600

The HTR-PM600 high-temperature gas-cooled reactor nuclear power plant is based on the technology of the high-temperature gas-cooled reactor pebble-bed module (HTR-PM) demonstration project. It utilizes proven HTR-PM reactor and steam generator modules with a thermal power of 250 MWth and power generation of approximately 100 MWe per module. Six modules in parallel, connected to a steam turbine, form a 600-MWe nuclear power plant. In addition, its system configuration in the nuclear island is identical to that of the HTR-PM in which the technical risks are minimized. Under this principle, the HTR-PM600 achieves the same level of inherent safety as the HTR-PM. The concept of a ventilated low-pressure containment (VLPC) is unchanged; however, a large circular VLPC accommodating all six reactor modules is adopted rather than the previous small-cavity-type VLPC, which contains only one module, as defined for the HTR-PM. The layout of the nuclear island and its associated systems refer to single-unit pressurized water reactor (PWR) practices. With this layout, the HTR-PM600 achieves a volume size of the nuclear island that is comparable to a domestic PWR of the same power level. This will be a Generation IV nuclear energy technology that is economically competitive.

Experimental analysis of a large range of the load capacity for the axial AMB in the HTR-PM

An axial active magnetic bearing (AMB) is equipped to provide axial support of the vertical rotor of the helium circulator in the High Temperature Gas-cooled Reactor Pebble Bed Module (HTR-PM). The axial AMB is required to support the rotor and the aerodynamic load. High-strength alloy steel is selected as the material of the thrust disk. Maximum load capacity of the axial AMB in the HTR-PM helium circulator is analyzed with the consideration of the rotating velocity and the material strength. Also, experiments are conducted based on a 1:1 prototype to reveal a large range of the load capacity for the axial AMB. The dynamic electromagnetic force corresponding to the allowable stress, the rotational velocity, and other parameters of the axial AMB, such as B-R Curve, resistance and inductance, are revealed. Effect of the heat treatment to the load capacity is also discussed. The results will provide some theoretical and engineering reference to the design and the type selection of the axial AMB in the rotor system of the HTGR (High Temperature Gas-cooled Reactor) energy conversion system.

Design, Experiment, and Commissioning of the Spent Fuel Conveying and Loading System of HTR-PM

The Chinese high-temperature gas-cooled reactor pebble-bed module, HTR-PM, began fuel loading in August 2021. The reactor refuels continuously, while the spent fuel is discharged from the core. The spent fuel conveying and loading system was designed to convey the spent fuel pebbles to the spent fuel building and load them into dry canisters for on-site interim storage. This study describes the operating principles of the main functions and introduces the experiments and commissioning tests of the system. Functional tests were carried out to indicate the items of mechanical and electrical equipment are functioning in accordance with the designed requirements. Experience learned from commissioning activities was also presented as feedback for future operation and design improvement.

Design optimization on characteristics of packed-bed thermal energy storage system coupled with high temperature gas-cooled reactor pebble-bed module

The packed-bed thermal energy storage (PBTES) coupled with the high temperature gas-cooled reactor pebble-bed module (HTR-PM) system is of emerging interest which can ensure the economy, safety and reliability operation of the HTR-PM. For the coupled PBTES system, multi-parameters need to be considered and there is a need to propose an integral method to help the design and optimization. In this work, firstly, the numerical results are compared and validated with the experiment data. Then, the thermal characteristics of the coupled PBTES system including the overall heat storage quantity, heat release quantity, average charging power, average discharging power, charging/discharging/total enthalpy efficiency and charging/discharging/total exergy efficiency are systematically evaluated with the orthogonal design optimization. At last, the typical non-dimensional numbers including the ReP number, Nu number and Pr number of the high temperature fluid (HTF) are illustrated for the optimal scheme. The results indicate that the calculation results agree well (relative error within 7.0%) with the experiment data. Additionally, the most optimal scheme is determined as 26 mm of the capsule diameter, “K2CO3-Li2CO3 (65–35 wt%)”+“LiF-NaF-KF (29–12–59 wt%)” of the phase change materials (PCMs), 0.229 m3·s−1 of the inlet HTF flow rate and 850℃ of the inlet HTF temperature or initial PCMs temperature. It demonstrates an integral method including the numerical model calculation, orthogonal design optimization, thermal characteristics evaluation, range and variance analysis, enthalpy and exergy efficiency and dimensionless number for the design of the coupled PBTES system.

Activation Analysis and Evaluation of Radionuclide Inventory Decay Heat for EU DEMO Vacuum Vessel Components

Activation inventory data were assessed for different European DEMOnstration Fusion Power Reactor (EU DEMO) vacuum vessel (VV) segments in order to evaluate the activity, dose rates, and decay heat of the structural materials and the impact of heterogeneity as a part of the Safety and Environment project of the EUROfusion/Power Plant Physics Technology program. Activity is a good indicator of the condition of irradiated material while decay heat and dose rate estimates are crucial for power plant operation and public safety. In addition, the VV structures were analyzed in terms of different breeder blanket configurations that were implemented in the model. Helium-cooled pebble bed and water-cooled lithium-lead breeder blanket modules were considered. These modules directly shield the VV from neutron irradiation; therefore, their design has a great influence on subsequent structure activation.

Study on the representativeness of airborne effluent sampling in the stack of a high-temperature gas-cooled pebble-bed modular reactor

Monitoring the airborne radioactive effluent from the stacks of nuclear power plants is very important. In order to obtain reliable and accurate measurements, the samples acquired from the stack must be representative of the whole. Thus, sampling should be performed at a location where the airflow is well mixed. We studied the transport behavior of airborne radioactive substances in the stack of a high-temperature gas-cooled reactor pebble-bed module (HTR-PM) demonstration power plant using a computational fluid dynamics method. The distributions of effluent flow velocity and concentration have been determined and compared to the mixing uniformity criteria of the ISO-2889 standard, then the source of uniformity is analyzed from vorticity distribution, particle trajectory, cyclone angle and coefficient of variation. We have determined the well-mixed location within the HTR-PM stack where a representative sample can be obtained. We discussed methods for the improvement of flow mixing and analyze the key factors involved.

Prediction calculations for the first criticality of the HTR-PM using the PANGU code

The high-temperature reactor pebble-bed module (HTR-PM) is a modular high-temperature gas-cooled reactor demonstration power plant. Its first criticality experiment is scheduled for the latter half of 2021. Before performing the first criticality experiment, a prediction calculation was performed using PANGU code. This paper presents the calculation details for predicting the HTR-PM first criticality using PANGU, including the input model and parameters, numerical results, and uncertainty analysis. The accuracy of the PANGU code was demonstrated by comparing it with the high-fidelity Monte Carlo solution, using the same input configurations. It should be noted that keff can be significantly affected by uncertainties in nuclear data and certain input parameters, making the criticality calculation challenge. Finally, the PANGU is used to predict the critical loading height of the HTR-PM first criticality under design conditions, which will be evaluated in the upcoming experiment later this year.

Thermodynamic analysis and optimization for steam methane reforming hydrogen production system using high temperature gas-cooled reactor pebble-bed module

ABSTRACT Thermodynamic analysis and optimization for the steam methane reforming (SMR) hydrogen production system using the high temperature gas-cooled reactor pebble-bed module power plant (HTR-PM) are investigated in this work. Based on the thermodynamic-equilibrium model, parameters of thermal efficiency (η), hydrogen production (Y H2/methane), methane conversion rate (X 1) and carbon monoxide conversion rate (X 2) are calculated under the specified temperature (T), pressure (P) and water-to-carbon ratio (S). The influence of S, T, P on η, Y H2/methane, X 1 and X 2 is then analyzed. It considers a wide range of operating conditions (T = 400–1200°C; P = 2–7 MPa and S = 2–10). The results show that the influence of on the system performance is significant. When T > 950°C, η and Y H2/methane increases slowly (4 < S ≤ 6) or reduces (S > 6). For the operating conditions of HTR-PM (P = 7 MPa; S = 6 and T = 950°C), the maximum value of η is 63.44% and the maximum Y H2/methane is 3.3 mol. At last, system optimized parameters are illustrated.

Experimental research on vertical mechanical performance of embedded through-penetrating steel-concrete composite joint in high-temperature gas-cooled reactor pebble-bed module

The high-temperature gas-cooled reactor pebble-bed module project is the first commercial Generation-IV NPP(Nuclear Power Plant) in China. A new joint is used for the vertical support of RPV(Reactor Pressure Vessel). The steel corbel is integrally embedded into the reactor-cabin wall through eight asymmetrically arranged pre-stressed high-strength bolts, achieving the different path transmission of shear force and moment. The vertical monotonic loading test of two specimens is conducted. The results show that the failure mode of the joint is bolt fracture. There is no prominent yield stage in the whole loading process. The stress of bolts is linearly distributed along the height of corbel at initial loading. As the load increases, the height of neutral axis of bolts gradually decreases. The upper and lower edges of the wall opening contact the corbel plate to restrict the rotation of the corbel. During the loading, the pre-stress of some bolts decreases. The increase of the pre-stress strength ratio of bolts has no noticeable effect on the structure stiffness, but it reduces the ultimate bearing capacity of the joint. A simplified calculation model for the elastic stage of the joint is established, and the estimation results are in good agreement with the experimental results.

Transient Study on the HTR-PM with TINTE-vPower Coupling Code Package

The high-temperature gas-cooled reactor pebble-bed module (HTR-PM) nuclear power plant consists of two nuclear steam supply system modules, each of which drives the steam turbine by the superheated steam flow and is fed by the heated-up water flow. The shared steam/water system induces mutual effects on normal operation conditions and transients of the nuclear power plant, which is worthy of safety concerns and intensive study. In this paper, a coupling code package was developed with the TINTE and vPower codes to understand how the HTR-PM operated. The TINTE code was used to analyze the reactor core and primary circuit, while the vPower code simulated the steam/water flow in the conventional island. Two TINTE models were built and coupled to one vPower model through the data exchange in the steam generator models. Using this code package, two typical transients were simulated by decreasing the primary flow rate or introducing the negative reactivity of one module. Important parameters, including the reactor power, the fuel temperature, and the reactor inlet and outlet helium temperatures of two modules, had been studied. The calculation results preliminarily proved that this code package can be further used to evaluate working performance of the HTR-PM.