High Temperature Gas-cooled Reactor Pebble-bed Module Research Articles

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.

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.

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.

Exergy study of hydrogen cogeneration and seawater desalination coupled to the HTR-PM nuclear reactor

This work presents a novel integration system of the high-temperature gas-cooled reactor-pebble bed module project to a hydrogen production process using the iodine-sulfur cycle in cogeneration with seawater desalination. The current approach includes a Rankine cycle, a sulfur-iodine thermochemical cycle for hydrogen production and a multi-stage flash desalination process. The use of a catalyst that allows the H2SO4 decomposition reaction to being carried out at temperatures compatible with the nuclear reactor project is considered. The residual heat from the acid decomposition reactions is used to desalinate seawater through the multi-stage flash process. A chemical process simulator is used to create a computational model that allows estimates of global and local efficiencies of the proposed flow diagram. Some operating parameters were sized, and their influence on the efficiency is also reported. The proposed model for the sulfur-iodine cycle can produce 0.41 kg/s of hydrogen with partial energy and exergetic efficiency of 37.35% and 38.64%. The desalination process can process 40.70 kg/s with energy and exergy efficiencies of 58.78% and 82.66%, respectively. The higher exergy destruction share is obtained in the heat exchangers (36.55%), chemical reactors (16.56%) and separators (12.80%). The global system showed efficiencies of 40.13% and 52.04%, respectively.

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.

A plant operating state analysis method in probabilistic safety assessment for multi-unit nuclear power plant

The nuclear power plants (NPPs) with more than one unit have become widespread globally, and plant operating state (POS) is a key issue to understand and manage characteristics of NPPs, especially to the safety level. Probabilistic safety assessment (PSA) is an indispensable safety analysis tool for NPPs, while it almost addresses single unit in practice at present. To perform PSA for multi-unit NPPs, the definition and evaluation of POS related to all units should be presented as preconditions. In this paper, a general framework of multi-unit POS analysis method based on single-unit POS is proposed, and the corresponding algorithm is also implemented. A case study on high temperature gas-cooled reactor - pebble-bed module (HTR-PM) is realized to verify the feasibility of this method. All possible combinations of POSs and their risk profiles for HTR-PM are obtained which could provide guidance for the subsequent operation under low power and shutdown conditions.

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.

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.

Steady-state thermal fluids analysis for the HTR-PM equilibrium core

The high temperature gas-cooled reactor-pebble bed module (HTR-PM), designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University, is a demonstration project of nuclear power plant with modular high temperature reactors (HTRs) in China. So far the detailed engineering design of the HTR-PM has been completed, and now the plant is at its commissioning stage. In the design process of the reactor, the thermal hydraulics design is of extreme importance for the evaluation of the reactor performance under various operation conditions. In this study, the steady-state thermal fluids analysis is performed for the HTR-PM equilibrium core. The analysis methods are introduced and the calculating models are presented. The key thermal-fluid parameters under different power levels are obtained and evaluated, including the temperature field of the reactor, maximum temperatures of the fuel and other components, pressure loss of the core and the primary circuit, and so on. As the most important safety-related parameter, the maximum fuel temperature is much lower than its safety limit for normal operation. The analysis results show that the thermal hydraulics design provides sufficient cooling capability to the core and thus ensures the reactor safety under normal operation.

A New Precursor Integral Method for Solving Space-Dependent Kinetic Equations in Neutronic and Thermal-Hydraulic Coupling System

The accurate prediction of the neutronic and thermal-hydraulic coupling system transient behavior is important in nuclear reactor safety analysis, where a large-scale nonlinear coupling system with strong stiffness should be solved efficiently. In order to reduce the stiffness and huge computational cost in the coupling system, the high-performance numerical techniques for solving delayed neutron precursor equation are a key issue. In this work, a new precursor integral method with an exponential approximation is proposed and compared with widely used Taylor approximation-based precursor integral methods. The truncation errors of exponential approximation and Taylor approximation are analyzed and compared. Moreover, a time control technique is put forward which is based on flux exponential approximation. The procedure is tested in a 2D neutron kinetic benchmark and a simplified high-temperature gas-cooled reactor-pebble bed module (HTR-PM) multiphysics problem utilizing the efficient Jacobian-free Newton–Krylov method. Results show that selecting appropriate flux approximation in the precursor integral method can improve the efficiency and precision compared with the traditional method. The computation time is reduced to one-ninth in the HTR-PM model under the same accuracy when applying the exponential integral method with the time adaptive technique.

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.

Rapid accident source term estimation (RASTE) for nuclear emergency response in high temperature gas cooled reactor

Rapid accident source term estimation (RASTE) is one of the key functions required by nuclear power plant (NPP) emergency. The results of RASTE are provided to the nuclear decision-making system in emergency, response to ranking accident by INES table and determining suitable protection. This paper proposes the RASTE system for high temperature gas-cooled reactor (HTR) to estimate accident source term based on the simplified methodology used in safety analysis report (SAR) and more realistic operation condition before and in an accident. RASTE simplifies the expression for normal radioactivity in the primary loop, re-suspended radioactivity, additional radioactivity release during the heating process, graphite dust generation, desorbed radioactivity from corrosion of graphite matrix and radioactivity released from failed coated fuel particle. Modifications are made to correct the different ingress water amounts and different fuel burnup. Some unclearly known parameters and processes are estimated from the operation experience as well as accident analysis by conservation rule. The integrated code of fundamental framework and user interface are also developed. This paper takes the High Temperature Gas Cooled Reactor Pebble-bed Module (HTR-PM), a demonstration project being constructed at Shidao Bay, Shandong Province, China, as the study case to discuss the applicability of RASTE. Considering three typical HTR-PM design basis accidents (DBAs), this paper selects P2, P4 as the representatives of LOCA accidents, selects SG6 and SG91 as the representatives of ingress accidents and selects T2 as the representative of transient accidents. Accuracy verification is made by comparing the environmental release estimated by RASTE to that in SAR. These compared targets include nuclide groups of inert gas, iodine, and aerosols, and key nuclides of Xe-133, I-131, Cs-134. Calculation results show that the methodology presented and RASTE is fit to estimate accident of HTR-PM. The development of RASTE system will contribute to form adequate capability of rapid emergency response for HTR nuclear power plant.

Cascaded HTGR Power-Level Control Only by Regulating Primary Helium Flow Rate

Reactor power-level control is a key technique of enhancing the safety, stability, and efficiency of nuclear plants. Currently, nearly all the power-level control methods are based on the scheme of inducing reactivity directly via control rods. However, the reactivity can also be injected indirectly via the negative temperature feedback effect, which results in the opportunity to develop indirect nuclear reactor power-level control methods. Motivated by this, a cascaded power-level control of high-temperature gas-cooled reactor, which is composed of a helium temperature controller and a neutron flux controller, is proposed based on only regulating the primary helium flow rate. The temperature controller, which is in the inner loop, regulates the primary helium flow rate according to the setpoint of helium temperature. The flux controller, which is in the outer loop, revises the setpoint of the inner-loop controller so as to eliminate the error of reactor neutron flux. It is proved that the closed-loop system can be guaranteed to be globally asymptotically stable. Finally, numerical simulation results corresponding to the application to the high-temperature gas-cooled reactor pebble-bed module (HTR-PM) plant verifies the theoretical result while showing the satisfactory performance as well as the influence of the controller parameters.

Installation of the graphite internals in HTR-PM

Internals are one of key equipment in the nuclear islands of High Temperature gas-cooled Reactor Pebble-bed Module (HTR-PM). Their structural integration is very important to the nuclear safety. Different from the internals in pressurized water reactors, a large number of graphite and carbon components are used in the internals of HTR-PM. These components are connected by dowels and keys with gaps between them to compensate the deformation caused by thermal and irradiation stresses. High installation accuracy of each component is required for the important safety and function related devices installed inside the internals, such as control rod and fuel handling system. Due to the cumulative assembly error of the components and the deformation of the supporting structures, the installation of the graphite and carbon internals is very challenging. Detailed requirements for the examination and installation of the graphite component have been developed and continually optimized for both trial assembly and assembly in the reactor. Research was also carried out for measurement techniques, lifting tools, examination and rapid assessments. The installation of graphite and carbon internals has been successfully completed for both modules. Some major installation results, including the position of the channels and levelness of the graphite components as well as the considerations of their acceptance are introduced in the present paper. The lessons-learned and experience can be applied to the installation of internals in HTRs and other reactors using graphite components, such as molten salt reactor.

Discussion on the accident behavior and accident management of the HTGR

High Temperature Gas-cooled Reactor (HTGR), which plays an important role in the worldwide development of Generation-IV nuclear energy technology, is well-known for the high safety, high efficiency and process heat application. A commercial-scale 200 MWe High Temperature gas-cooled Reactor Pebble-bed Module (HTR-PM) has been designed and is now under construction in Shandong Province, China. Most of the construction and installation work have been finished and the connection to the electric grid will be expected in the end of 2020.In this paper, the reactor behaviors of the HTR-PM during the typical design basis accidents (DBAs) and beyond design basis accidents (BDBAs) have been introduced. In the DBAs, the maximum fuel temperature will never exceed its design limitation of 1620 °C, below which under the design burn-up the “Tristructural-isotropic” (TRISO) coated particles can effectively retain their fission-product inventory. Even in the typical BDBAs with extremely low probability, there is enough time, e.g. several days, to adopt appropriate measures to mitigate the consequence, so that the large release of the radioactive materials would not happen.Accident management is important for the nuclear power plant. Besides, After the Fukushima Dai-ichi nuclear accident, in the worldwide, people are more concerned about the severe accident management of the nuclear power plant, and some standards and guidelines are issued. Based on the accident analyses of the HTR-PM, its accident management is studied and discussed. The designers believe that the accident management procedures can be simplified and no offsite emergency measures are needed for the HTR-PM due to the inherent safety design. Compared to the other types of the nuclear power plant, e.g. the Pressurized-Water Reactor (PWR) power plant, the Severe Accident Management Guideline (SAMG) can be simplified or even unnecessary for the HTR-PM.Above work also can provide reference for the further study on the safety regulations, standards or guidelines for the design, operation and accident management of the HTGR.