Temperature Gas-cooled Reactor Pebble-bed Module Research Articles

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.

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.

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.

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.

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.

Frequency domain analysis of two-phase flow instabilities in a helical tube once through steam generator for HTGR

Two-phase flow instabilities will disturb the control systems and cause mechanical vibrations and failure of components, which should be avoided. Two-phase flow instabilities in a helical tube Once Through Steam Generator (OTSG) for High Temperature Gas-cooled Reactor (HTGR) were investigated. First, a linear frequency domain model of the helical tube OTSG was developed. The convective evaporation process was divided into single phase water region, two-phase flow boiling region and superheated steam region. Incompressible assumptions were adopted in the single phase water and superheated steam regions. Homogeneous flow model was used in the two-phase region. Small perturbation linearization and Laplace transform were performed to analyze the governing equations in the three regions. The corresponding frequency domain transfer functions describing the flow pulsation and pressure drop pulsation of the two-phase flow systems were derived. By solving the Nyquist curve of the transfer function with matlab software, the two-phase flow instabilities and the stable boundary of the evaporation process in helical tubes can be determined. The new model in this paper was verified using previous experimental results. The two-phase flow stability in the helical tube OTSG of High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) was analyzed using the developed frequency domain model at the design power levels. The stability boundary of the two-phase flow and effects of various parameters on flow instabilities were also investigated. Then an accumulator was added to this linear frequency domain model, which was used to study the effect of compressible volume or accumulator on DWO and PDO. Using this model, the effects of throttle valves at different locations on the stability of the system with an accumulator were also studied. The results show that the helical tube OTSG for HTR-PM can operate stably under the designed working conditions with enough margin.

Thermal analysis of HTGR helical tube once through steam generators using 1D and 2D methods

The working temperature of High Temperature Gas-cooled Reactor (HTGR) steam generators is much higher than that of Pressurized Water Reactor (PWR) steam generators. Large temperature non-uniformities will make the HTGR steam generator deviate from the designed working temperatures, which should be avoided. The temperature distribution and thermal deviation should be analyzed. One-dimensional (1D) and two-dimensional (2D) methods for analyzing the thermal hydraulic performance of an HTGR steam generator were developed. In the 1D method, the tube and shell side governing equations were solved. They were coupled through the energy source terms. In the 2D method, simulation of the shell side helium cross flow over tube bundles was based on the two-dimensional CFD code Ansys Fluent. The tube side water flow was coupled to the 2D simulation of the cross flow over tube bundles through the heat transfer between the primary and secondary sides. The 1D and 2D methods were validated using the measured data of the Thorium High Temperature Reactor (THTR) steam generators and the High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) test steam generator. The 1D method can give the overall thermal performance of the steam generators. The 2D method can predict temperature non-uniformities in the tube bundles. The overall thermal performance of the steam generator was simulated. The thermal non-uniformities or thermal deviations caused by geometrical and thermal hydraulic deviations were analyzed. Radiation heat transfer contributes to 0.9% of the total thermal power. A one percent change of the secondary side water flow rate of a certain layer of tubes will result in a 5–6 °C variation of the outlet steam temperature of this layer of tubes. A helical diameter deviation of 1 mm will result in a maximum outlet steam temperature variance of 6–7 °C. An inlet helium temperature profile with a 10 °C variance will only result in a variation of 2 °C in the outlet steam temperature.

Supercritical steam generator design and thermal analysis based on HTR-PM

The high temperature gas-cooled reactor pebble-bed module (HTR-PM) being built by the Institute of Nuclear and New Energy Technology of Tsinghua University in China is a typical Generation IV nuclear power system. The subcritical steam generators in HTR-PM can be replaced by the supercritical steam generators to work with the current reactors and a supercritical steam turbine to improve the thermal efficiency, which is being presented in the third technical phase of the HTR-PM program in China. The supercritical high temperature gas-cooled reactor pebble-bed module system scheme is proposed based on HTR-PM. A supercritical steam generator with helically coiled tubes is designed. Parameters of the geometry and the design conditions are listed in detail. The heat transfer performance of the supercritical steam generator is analyzed, and the heat flux, the temperature variation and the heat transfer coefficient along the tube are shown. The effects of the transverse and longitudinal pitches between the tubes in the supercritical steam generator are very important. The tube length, tube bundle height, and the deviation of the tube length in various layers are compared with various transverse and longitudinal pitches and the optimal relative transverse and longitudinal pitches are illustrated.

Effects of ion irradiation and oxidation on point defects in IG-110 nuclear grade graphite

Nuclear grade graphite is a kind of key material in the high temperature gas-cooled reactor pebble-bed module (HTR-PM), where nuclear grade graphite acts as the fuel element matrix material, structural material and neutron reflector. In the reactor, the service environment of nuclear grade graphite suffers high temperature and strong neutron radiation. Both neutron radiation and the oxidation by the oxidizing impurities in HTGR coolant can cause the structure to damage and the properties to deteriorate. Therefore, it is of great significance to study the evolution of defects in nuclear grade graphite for improving the reactor safety. The effects of ion irradiation and oxidation on the point defects in IG-110 graphite are studied in this work. The 190 keV He&lt;sup&gt;+&lt;/sup&gt; implantation treatments at room temperature with fluences of 1 × 10&lt;sup&gt;15&lt;/sup&gt;, 5 × 10&lt;sup&gt;15&lt;/sup&gt;, 1 × 10&lt;sup&gt;16&lt;/sup&gt; and 1 × 10&lt;sup&gt;17&lt;/sup&gt; cm&lt;sup&gt;–2&lt;/sup&gt; are performed to induce 0.029, 0.14, 0.29 and 2.9 displacements per atom respectively. Oxidation treatments are performed at 850 ℃ for 10, 15, 20 and 25 min. Different sequences of He&lt;sup&gt;+&lt;/sup&gt; ion irradiation and oxidation are performed, which include irradiation only (Irr.), oxidation only (Ox.), irradiation followed by oxidation (Irr.-Ox.), and oxidation followed by irradiation (Ox.-Irr.). Raman spectrum shows that with the increase of ion irradiation dose, the intensity ratio of D peak to G peak (&lt;i&gt;I&lt;/i&gt;&lt;sub&gt;D&lt;/sub&gt;/&lt;i&gt;I&lt;/i&gt;&lt;sub&gt;G&lt;/sub&gt;) first increases and then decreases, implying that the point defects in graphite are induced by ion irradiation and the point defects evolve as dose increases; the degree of graphitization increases after oxidation, implying that the point defects are recovered by the annealing effect at high temperature, and the point defects decrease after oxidation. This makes Ox.-Irr. samples have a lower point defect content than Irr. samples, and leads Irr.-Ox. samples to possess a higher point defect content than Ox. samples. The positron annihilation Doppler broadening tests reveal that there are only point defects after ion irradiation and oxidation have partially recovered point defects. The ion irradiation and oxidation have opposite effects on the evolution of point defect in graphite. The ion irradiation increases the average &lt;i&gt;S&lt;/i&gt;-parameter and reduces the average &lt;i&gt;W&lt;/i&gt;-parameter, while oxidation reduces the average &lt;i&gt;S&lt;/i&gt;-parameter and increases the average &lt;i&gt;W&lt;/i&gt;-parameter. The annealing effect at 850 ℃ cannot completely recover the point defects in Irr.-Ox. samples.

Influences of tube wall on the heat transfer and flow instability of various supercritical pressure fluids in a vertical tube

The subcritical steam generators in the current high temperature gas-cooled reactor pebble-bed module, HTR-PM, in China can be replaced by supercritical steam generators to better coordinate the reactors and the supercritical steam turbine unit to improve the thermal efficiency with no fluid phase change at supercritical pressures. This study then numerically analyzes the heat transfer and flow instabilities of supercritical pressure fluid flow in a vertical tube with zero or finite-thickness walls. The wall thickness has negligible effect on the heat transfer at supercritical pressure water at steady state. However, for transient calculations, the flow rate oscillates intensely beyond a critical heat flux with the zero-thickness wall, while the flows in tubes with finite-thickness walls are very different. The wall properties including density, specific heat and thermal conductivity are varied to analyze why the oscillations are suppressed by the finite-thickness wall. The flow and heat transfer characteristics and the heat stored in the wall compared with the total power are presented during the heating process at various moments. The flow instabilities are also analyzed for water at various pressures and CO2 to analyze the effects of the fluid properties.

Preliminary phenomena identification and ranking tables on the subject of the High Temperature Gas-cooled Reactor-Pebble Bed Module thermal fluids and accident analysis

The High Temperature Gas-cooled Reactor-Pebble Bed Module (HTR-PM) is a demonstration power plant designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University of China. In conformance with China’s nuclear safety requirements related to the development, assessment, application and quality assurance of computer software used in the safety analysis of nuclear power plants (NPPs), a phenomena identification and ranking table (PIRT) process was conducted on the subject of the HTR-PM thermal fluids and accident analysis by an expert panel sponsored by INET, on the basis of the latest HTR-PM design scheme. The PIRT objective is to identify safety-relevant phenomena that may occur during normal operation and under accident conditions, and to assess research activities pertaining to code development and assessment, such as the code validation in INET.In the PIRT process, safety-related phenomena were identified by the expert panel for the following HTR-PM scenarios: normal operation (NO), pressurized loss of forced cooling (PLOFC), depressurized loss of forced cooling (DLOFC), air ingress following DLOFC, steam/water ingress and anticipated transients without scram (ATWS). Subsequently, the relative importance of each phenomenon was evaluated according to its effect or influence on the figures of merit established by the expert panel, such as the fuel failure fraction and the maximum fuel temperature. In addition, each identified phenomenon was assigned a knowledge level rank by the expert panel, indicating the prediction capability for this phenomenon based on existing analytical tools and test data. What is more, all of the PIRT elements were well documented by a special report, including PIRT issues, objectives, descriptions of the HTR-PM plant and accident scenarios, evaluation criteria, current knowledge base, ranking scales of relative importance and knowledge level, a set of PIRTs, and expert discussions.

Numerical simulation on heat transfer process in the reactor cavity of modular high temperature gas-cooled reactor

In order to enhance the safety of the reactor, the passive residual heat removal system (PRHRS) is adopted by both the 10MW High Temperature Gas-cooled Test Reactor (HTR-10) and the High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). The heat transfer process in the reactor cavity is very important for the whole PRHRS to remove the heat from the reactor, especially once an accident occurs. A three-dimensional numerical model used to simulate the heat transfer process in the reactor cavity of HTR-10 was developed in this study, in which the flow of water in the water cooled pipe was included. The effects of different symmetry boundary conditions on the heat transfer process were studied. Three experimental conditions of HTR-10 were simulated, and the local and global heat transfer processes were analyzed. The results show that the effects of symmetry boundary condition can be neglected when two or more water cooled pipes are included in the model. The temperature of air varies greatly near the reactor pressure vessel (RPV) and the water cooled panel. The maximum temperature occurs at the top of the RPV. With the elevation increasing, the rising rate of water temperature gradually increases. Moreover, radiation become more dominant as the RPV temperature rises.

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.

Design and Development Framework of Safety-Critical Software in HTR-PM

With the development of information technology, the instrumentation and control system of nuclear power plant nowadays rely heavily on the massive and complex software to ensure the safe and efficient operation of the power plant. The improvement of the software design and development for the safety systems has been a research focus for its decisive impact on the nuclear safety. The framework of the software design and development for reactor protection system in High Temperature Gas-Cooled Reactor-Pebble bed Module was introduced in this paper. Firstly, during the design period, in addition to multichannel redundancy, grouping of protection variables and diverse 2-out-of-4 logics were adopted by different subsystems of each channel in case of common cause failure. Then a series of development characteristics together with strict software verification and validation were performed. Thirdly, during the software test period, an improved software reliability growth model based on the Goel-Okumoto model according to the analysis of fault severity was proposed to help in estimating the reliability of the software product and identifying the software release time.

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.

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.

Carbide evolution and service life of simulated post weld heat treated 2.25Cr–1Mo steel

Abstract To estimate the service life of the reactor internals made of 2.25Cr–1Mo steel in the high temperature gas-cooled reactor pebble-bed module demonstration plant, creep rupture test was performed on the simulated post weld heat treated steel at 773 K over a range of stress levels 300–200 MPa. It has been shown that the rupture time is 126–5532 h, correspondingly. Then, the service life under stress of 161.2 MPa at 648 K was calculated to be so long as 210,374,853 h (≈24,015 years). Four types of carbides M 3 C, M 2 C, M 7 C 3 and M 23 C 6 existed in both the steel plate and the post-rupture specimens. As the rupture time increases, the weight percentages of M 23 C 6 and M 2 C that are stable in thermodynamics in the post-rupture specimens increase while those of metastable M 3 C and M 7 C 3 decrease on the whole. During the long term creep exposure, the deformation promotes the precipitation of M 3 C, M 2 C and M 23 C 6 , according to the experimental results with the deformed and undeformed parts of the post-rupture specimens, but its influence on the weight percentage of M 7 C 3 seems rather complicated.

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.

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.