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

Theoretical Analysis of Effective Thermal Conductivity for the Chinese HTR-PM Heat Transfer Test Facility

The Chinese high temperature gas-cooled reactor pebble bed module (HTR-PM) demonstration project has attracted increasing attention. In order to support the project, a large-scale heat transfer test facility has been constructed for pebble bed effective thermal conductivity measurement over the whole temperature range (0~1600 °C). Based on different heat transfer mechanisms in the randomly packed pebble bed, three different types of effective thermal conductivity have been theoretically evaluated. A prediction of the total effective thermal conductivity of the pebble bed over the whole temperature range is provided for the optimization of the test facility and guidance of further experiments.

Source Term Study on Tritium in HTR-PM: Theoretical Calculations and Experimental Design

The high temperature gas-cooled reactor pebble-bed module (HTR-PM) in China received much attention for its inherent safety performance and high thermal efficiency. The generation mechanism, distribution, reduction route, and release type of tritium (H-3) in HTR-PM are presented with a complete theoretical model. The calculation results indicate the majority of H-3 in the core is generated by the activation reaction of B-10. The activity concentration of H-3 in the primary loop and the specific activity of H-3 in the secondary loop at the operating equilibrium are computed as 3.69 × 106 Bq/m3STPof helium and 4.22 × 104 Bq/kg of water, respectively. The H-3 sampling measurement in HTR-PM has been designed to collect data from the primary coolant, from the liquid waste storage tank, from the secondary coolant, and from the liquid and gaseous effluents, separately. In this paper, the design of H-3 sampling positions in the helium purification system is discussed. The H-3 sampling measurement from the primary helium in HTR-PM has been improved, which can provide reliable activity concentration data of H-3 in the primary loop and supply accurate evaluation for the efficiency of the helium purification system.

An integrated modeling approach for event sequence development in multi-unit probabilistic risk assessment

To obtain a complete site risk profile for multi-unit nuclear power plants (NPPs), the existing single-unit probabilistic risk assessment (PRA) model needs to be upgraded to address both intra-unit and inter-unit dependencies. The high temperature gas cooled reactor-pebble bed module (HTR-PM), with two reactor-steam generator modules and one shared steam turbine-generator set, is a ready-in-hand pilot for multi-unit PRA (MUPRA) studies. In this paper, a general framework for MUPRA is firstly developed based on HTR-PM PRA methodology. An Integrated Modeling (IM) approach is then proposed for multi-unit event sequence modeling. The approach is discussed in detail and illustrated with a loss of offsite power (LOOP) case study. Also, the limitations of the current analysis are put forward for open discussion.

The Shandong Shidao Bay 200 MWe High-Temperature Gas-Cooled Reactor Pebble-Bed Module (HTR-PM) Demonstration Power Plant: An Engineering and Technological Innovation

ABSTRACT After the first concrete was poured on December 9, 2012 at the Shidao Bay site in Rongcheng, Shandong Province, China, the construction of the reactor building for the world's first high-temperature gas-cooled reactor pebble-bed module (HTR-PM) demonstration power plant was completed in June, 2015. Installation of the main equipment then began, and the power plant is currently progressing well toward connecting to the grid at the end of 2017. The thermal power of a single HTR-PM reactor module is 250 MW th , the helium temperatures at the reactor core inlet/outlet are 250/750 °C, and a steam of 13.25 MPa/567 °C is produced at the steam generator outlet. Two HTR-PM reactor modules are connected to a steam turbine to form a 210 MW e nuclear power plant. Due to China's industrial capability, we were able to overcome great difficulties, manufacture first-of-a-kind equipment, and realize series major technological innovations. We have achieved successful results in many aspects, including planning and implementing RD these obtained experiences may also be referenced by the global nuclear community.

Oxidation Analyses of Massive Air Ingress Accident of HTR-PM

The double-ended guillotine break (DEGB) of the horizontal coaxial gas duct accident is a serious air ingress accident of the high temperature gas-cooled reactor pebble-bed module (HTR-PM). Because the graphite is widely used as the structure material and the fuel element matrix of HTR-PM, the oxidation analyses of this severe air ingress accident have got enough attention in the safety analyses of the HTR-PM. The DEGB of the horizontal coaxial gas duct accident is calculated by using the TINTE code in this paper. The results show that the maximum local oxidation of the matrix graphite of spherical fuel elements in the core will firstly reach3.75⁎104 mol/m3at about 120 h, which means that only the outer 5 mm fuel-free zone of matrix graphite will be oxidized out. Even at 150 h, the maximum local weight loss ratio of the nuclear grade graphite in the bottom reflectors is only 0.26. Besides, there is enough time to carry out some countermeasures to stop the air ingress during several days. Therefore, the nuclear grade graphite of the bottom reflectors will not be fractured in the DEGB of the horizontal coaxial gas duct accident and the integrity of the HTR-PM can be guaranteed.

Experimental study on the adsorption and desorption of tritium in the graphite materials for HTR-PM

Tritium behavior in the reactor such as production, diffusion and release are accompanied by their adsorption and desorption in graphite materials, which are essential to the safety of high temperature gas cooled reactor (HTGR). In order to study this important issue, hydrogen instead of tritium is experimentally used in this work and justified viable by theory. By performing multiple sets of comparative experiments, the features of hydrogen adsorption and desorption behavior changing by adsorption temperature and time in typical graphites used in HTR-PM (High Temperature Gas Cooled Reactor – Pebble Bed Module), i.e. reflective layer, fuel element and boron carbon bricks, have been observed and analyzed. Furthermore, the adsorption rates of hydrogen in the three materials as above at different conditions are also given. Based on the experimental results, tritium behavior in the HTR-PM was inferred and estimated, which is significant for the further study on the mechanism of tritium transport.

An optimized process for tritium-containing waste water collection of High-Temperature Gas-cooled Reactor

An optimized process for tritium-containing waste water collection of High-Temperature Gas-cooled Reactor (HTGR) was developed and experimentally verified using the 10MW High-Temperature Gas-cooled Reactor-test module (HTR-10). Compared with the previous process, an auxiliary molecular sieve bed was added in helium purification regeneration system and new operation process was proposed to collect tritium-containing waste water. In this paper, the optimized process and verification experiment were presented in detail. In commissioning experiment of the improved HTR-10, a large quantity of high-dose tritium-containing waste water was successfully collected in the water separator of helium purification regeneration system, with the specific activity being 6.1×109Bq/L. The verification experiment confirms that the optimized process is effective and reliable for the demonstration plant design of High Temperature Gas-cooled Reactor-Pebble bed module (HTR-PM) to avoid the large emission of detrimentally radioactive waste water to the environment.

A design of human–machine interface for the two-modular high-temperature gas-cooled reactor nuclear power plant

A design of human–machine interface (HMI) has been developed to ensure the safety and availability of the two-modular high-temperature gas-cooled reactor nuclear power plant in China. As the first design for the plant with two modules coupled to one steam turbine, the staffing arrangement and HMIs are different from current nuclear power plants (NPPs). One control room is used to monitor and control the two reactor modules, the turbine generator, and the balance-of-plant (BOP). Considering the digital control room style, the elements of the HMIs in the control room are the large display panels and control consoles. According to the function analysis, the HMIs can be divided into three continuous parts: HMIs of reactor #1, conventional island, and reactor #2, from left to right for three operators respectively, and HMIs for the two reactor modules are basically the same. A Distributed control system (DCS) is employed for the non-safety controls, and computerized HMIs on the consoles are the interfaces for frequent use. The necessary safety operations and displays are set on the safety console of each reactor, independently. The large display panels are designed as an overview of the whole plant, which consist of large screen displays, mimic displays and alarm tiles. The failure modes and the diversities of HMIs are analyzed. Some special human factor engineering (HFE) strategies are considered for the control authorities of the two modules, the control of the systems shared between the two modules, etc. Two different solutions for the arrangement of HMIs and staffing are discussed. For lack of operating experiences, a verification platform is built to make verification and validation (V&V) of the HMIs. After sufficient validation, the design will be realized in the demonstration construction of the high-temperature gas-cooled reactor-pebble bed module (HTR-PM) in China.

Technical design and engineering prototype experiment of active magnetic bearing for helium blower of HTR-PM

High-temperature gas-cooled reactor-pebble bed module (HTR-PM) of China, based on the technology and experience of the 10MW high-temperature gas-cooled reactor (HTR-10), is currently in the construction phase. The helium blower is the key equipment in the primary loop of the HTR-PM. It drives the helium to flow in the primary loop for energy exchange. Active magnetic bearing (AMB) is replacing ordinary mechanical bearings as the perfect sustaining assembly for the helium blower because they have several advantages: they are free of contact, do not require lubrication, are not subject to the contamination of wear, have endurance, and control performance very well. The AMB is the important technology for the helium blower. The axial and radial AMB must have enough lifting capacity and reliability to support the rotor. In this paper, the structure and system characteristics of the helium blower will be discussed. The electromagnetic field analysis and the structure design of AMB will be described in detail. And the engineering prototype of helium blower with the AMB has been produced. Many experiments based on the engineering prototype with the AMB are being carried out, and the experiment’s scheme and plan will also be explained in this paper. These results will offer the important theoretical base and experience in the engineering application for the AMB design of the helium blower of HTR-PM.

Thermal hydraulic simulation of reactor of HTR-PM based on thermal-fluid network and SIMPLE algorithm

This paper provides the model development and its verifications for the reactor thermal-hydraulic transient model for the High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). A thermal-fluid network is constructed to simulate the complexity of the flow and heat transfer structure in the reactor of HTR-PM. SIMPLE algorithm is applied to solve the conservation equations of the thermal-fluid network to simulate the transient behavior of the high speed turbulent helium flow. A calculation process is proposed for coupling the high speed helium flow and the complex heat transfer structure. A FORTRAN code was developed based on the solution method and the thermal-fluid network. Several test cases including two steady states and a Control Rod Withdrawal accident are simulated by this code and the results are compared to those obtained by a safety analysis code, namely THERMIX. The good agreement between the two codes indicates that the proposed model and solution method based on SIMPLE algorithm is reasonable and applicable for simulating the thermal-hydraulic behavior in reactor of HTR-PM.

Preparation of ammonium diuranate particles by external gelation process of uranium in INET

In order to satisfy the mass spherical fuel elements (SFE) requirements for the High Temperature gas-cooled Reactor (HTR)–Pebble bed Module (HTR-PM), a modified and optimized fabrication of dense uranium dioxide (UO2) kernel via an acid-deficiency uranium nitrate (ADUN) solution was researched and developed in the institute of nuclear and new energy technology (INET). In this study, external gelation process of uranium (EGU) based on the total gelation of uranium (TGU) in the period of 10MW HTR (HTR-10) was used to prepare the ammonium diuranate (ADU) compound particles, an intermediate for UO2 kernel. The scale of batch production was improved from 500g U/batch at that time to current 3kg U/batch. Improvements in many aspects such as large scale preparation of ADUN, multiple-nozzle casting system and new aging, washing and drying (AWD) system without isopropanol were also presented. Under strict controls and measures, ADU particles with good sphericity and uniform diameter were acquired to ensure the subsequent preparation of dense UO2 kernels for HTR-PM.

Thermal hydraulic analysis of a pebble-bed modular high temperature gas-cooled reactor with ATTICA3D and THERMIX codes

Abstract Along with the further ongoing design procedures of a large demonstration power plant of Chinese 200 MWe high temperature gas-cooled reactor pebble-bed module (HTR-PM), three-dimensional (3-D) analysis is necessary for studying the reactor characteristics under some special unsymmetrical conditions. Compared with the well established two-dimensional thermal-hydraulics code THERMIX, a new three-dimensional system analysis code ATTICA3D (Advanced Thermal hydraulics Tool for In-vessel & Core Analysis in 3D), developed by IKE of University of Stuttgart on the basis of TH3D, is used to perform the steady state calculation and transient analysis of a depressurized loss of coolant accident (DLOCA) for the HTR-PM. After a brief introduction of the reactor primary circuit, the above two codes, including correlative typical parameters and relevant formulae are described in detail. Then an ATTICA3D model, which keeps consistent with the THERMIX model as much as possible, is constructed according to the preliminary design of the HTR-PM. Based on the cylindrically symmetrical power distribution, the comparisons of the two-dimensional calculations between ATTICA3D and THERMIX indicate, that the main results including pressure drop and temperature distribution show good agreement. Furthermore, considering one set of small absorber spheres (SAS) dropping into the side reflectors resulting in an asymmetric distribution of the nuclear power, a three-dimensional calculation for the steady state is implemented to basically show the good 3-D simulation capabilities of ATTICA3D, which will be helpful to the next refined design stage for the HTR-PM project. It should be pointed out that the current 3-D simulation results illustrated in this paper are still preliminary, and more research work, e.g., comparison with experimental data or with some commercial CFD codes, needs to be carried out in order to further validate the ATTICA3D code and study the special 3-D characteristics of the HTR-PM in the future.

Development of Probabilistic Safety Assessment with respect to the first demonstration nuclear power plant of high temperature gas cooled reactor in China

Due to the unique concept of HTR-PM (High Temperature Gas Cooled Reactor-Pebble Bed Module) design, Chinese nuclear authority has anticipated that HTR-PM will bring challenge to the present regulation. The pilot use of PSA (Probabilistic Safety Assessment) during HTR-PM design and safety review is deemed to be the necessary and efficient tool to tackle the problem, and is actively encouraged as indicated in the authority's specific policy statement on HTR-PM project. The paper summarizes the policy statement to set up the base of PSA development and application activities. The up-to-date status of HTR-PM PSA development and the risk-informed application activities are introduced in this paper as the follow-up response to the policy statement. For open discussion, the paper hereafter puts forward several technical issues which have been encountered during HTR-PM PSA development. Since HTR-PM PSA development experience has the general conclusion that many of the PSA elements can be and have been implemented successfully by the traditional PSA techniques, only the issues which extra innovative efforts may be needed are highlighted in this paper. They are safety goal and risk metrics, PSA modeling framework for the non-water reactors, passive system reliability evaluation, initiating events frequencies and component reliability data estimation techniques for the new reactors and so on. The paper presents the way in which the encountered technical issues were or will be solved, although the proposed way may not be the ultimate best solution. The paper intends to express the standpoint that although the PSA of new reactor has the inherent weakness due to the insufficient information and larger data uncertainty, the problem of component reliability data is much less severe than people have conceived. The unique design conception and functional features of the reactors can influence the results more significantly than the component reliability data. What we are benefited from PSA is indeed the systematic way which PSA follows. This is more important especially for the new reactors.

Simultaneous approach for simulation of a high-temperature gas-cooled reactor

The simulation of a high-temperature gas-cooled reactor pebble-bed module (HTR-PM) plant is discussed. This lumped parameter model has the form of a set differential algebraic equations (DAEs) that include stiff equations to model point neutron kinetics. The nested approach is the most common method to solve DAE, but this approach is very expensive and time-consuming due to inner iterations. This paper deals with an alternative approach in which a simultaneous solution method is used. The DAEs are discretized over a time horizon using collocation on finite elements, and Radau collocation points are applied. The resulting nonlinear algebraic equations can be solved by existing solvers. The discrete algorithm is discussed in detail; both accuracy and stability issues are considered. Finally, the simulation results are presented to validate the efficiency and accuracy of the simultaneous approach that takes much less time than the nested one.

Saturated Output Feedback Dissipation Steam Temperature Control for the OTSG of MHTGRs

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. The once through steam generator (OTSG) is a crucial element for any MHTGR power plants with steam cycle. In order to guarantee the efficient and safe operation of the power plant, the outlet steam temperature of the OTSG should be well controlled. In this paper, a saturated output feedback dissipation control (SOFDC) law is presented for regulating the outlet steam temperature of OTSGs, and the corresponding robustness analysis is also given. In the numerical study, the SOFDC is applied to the outlet steam temperature control for the OTSG of Chinese high temperature gas-cooled reactor pebble-bed module (HTR-PM) project. Numerical simulation results show not only feasibility but also good control performance of this regulator, and the influence of its parameters to the regulation performance is also discussed.

A nodal dynamic model for control system design and simulation of an MHTGR core

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high-temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high-temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power-distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.

Analysis of fuel element matrix graphite corrosion in HTR-PM for normal operating conditions

The corrosion of spherical fuel element by oxidizing gases will degrade their thermal and mechanical properties in pebble-bed reactors. Oxidizing impurities in primary helium coolant may have significant influence on the gasification of graphite during the long service time even though their concentrations remain relatively low. This paper deals with the corrosion of matrix graphite by steam and oxygen in Chinese high temperature gas-cooled reactor pebble-bed module (HTR-PM). The influence of steam contents was first analyzed, and then the effect of oxygen contents was also taken into account. The results show that the corrosion by steam was weak and it would not threaten the normal service of spherical fuel elements for expected steam contamination levels. On the contrary, the corrosion would be more severe while the oxygen content was as high as currently expected. Finally, the upper limits of steam and oxygen in primary helium coolant were recommended to be 1.0 and 0.05 cm 3 m −3.

Operation and Control Simulation of a Modular High Temperature Gas Cooled Reactor Nuclear Power Plant

Issues in the operation and control of the multi-modular nuclear power plant are complicated. The high temperature gas cooled reactor pebble-bed module (HTR-PM) plant with two-module will be built as a demonstration plant in China. To investigate the operation and control characteristics of the plant, a simplified dynamic model is developed and mathematically formulated based upon the fundamental conversation of mass, energy and momentum. The model is implemented in a personal computer to simulate the power increase process of the HTR-PM operation. The open loop operation with no controller is first simulated and the results show that the essential parameter steam temperature varies drastically with time, which is not allowable in the normal operation. According to the preliminary control strategy of the HTR-PM, a simple steam temperature controller is proposed. The controller is of Proportional-type with a time lag. The closed loop operation with a steam temperature controller is then implemented and the simulation results show that the steam temperature and also other parameters are all well controlled in the allowable range.

Economic potential of modular reactor nuclear power plants based on the Chinese HTR-PM project

Modular reactors with improved safety features have been developed after the Three-Mile Island accident. Economics of small modular reactors compared to large light water reactors whose power output is 10 times higher is the major issue for these kind of reactors to be introduced into the market. Based on the Chinese high temperature gas-cooled reactor pebble-bed module (HTR-PM) project, this paper analyzes economical potentials of modular reactor nuclear power plants. The reactor plant equipments are divided into 6 categories such as RPV and reactor internals, other NSSS components and so on. The economic impact of these equipments is analyzed. It is found that the major difference between an HTR-PM plant and a PWR is the capital costs of the RPV and the reactor internals. The fact, however, that RPV and reactor internals costs account for only 2% of the total plant costs in PWR plants demonstrates the limited influence of this difference. On the premise of multiple NSSS modules forming a nuclear power plant with a plant capacity equivalent to a typical PWR plant, an upper value and a target value of the total plant capital costs are estimated. A comparison is made for two design proposals of the Chinese HTR-PM project. It is estimated that the specific costs of a ready-to-build 2 × 250 MW th modular plant will be only 5% higher than the specific costs of one 458 MW th plant. When considering the technical uncertainties of the latter, a 2 × 250 MW th modular plant seems to be more attractive. Finally, four main points are listed for MHTGRs to achieve economic viability.

Thermal–hydraulic feasibility analysis on uprating the HTR–PM

In order to meet energy demand in China, the high temperature gas-cooled reactor–pebble-bed module (HTR–PM) is being developed. It adopts a two-zone core, in which graphite balls are loaded in the central zone and the outer part is fuel ball zone, and couple with a steam cycle. Outer diameter of the reactor core is 4.0 m and height of the core is 9.43 m. The helium inlet and outlet temperature are 250 and 750 °C, respectively. The reactor thermal power is 380 MW. Preliminary studies show that the HTR–PM is feasible technologically and economically. In order to increase the reactor thermal power of the HTR–PM, some efforts have been made. These include increasing the height of reactor core, optimizing the thickness of fuel zone and better selection of the scheme of central graphite zone, etc. Basic design concepts and thermal–hydraulic parameters of the HTR–PM are given. Measures to increase the thermal power are introduced. Thermal–hydraulic analysis results are presented. The results show that, from the viewpoint of thermal–hydraulics, it is possible to increase the reactor power.