With the continuous development of the nuclear power technology in the world, all countries in the world are becoming more and more interested in the inherent safety of nuclear power technology, while the research and development of the spherical bed type high temperature gas cooled reactor nuclear power technology in China has formally catered to this demand. As a major national science and technology project, since the construction of the high temperature gas cooled reactor demonstration project (HTR-PM) since 2012, the civil construction of the nuclear island has been basically completed, the installation of equipment has been carried out orderly, and many process systems have entered debugging and operation stage gradually. As an important auxiliary process system, fuel handling and storage system for online refueling of the pebble bed high temperature gas cooled reactor, plays an important role in relation to the stable operation of the reactor. The main functions of the fuel handling and storage system are loading the fresh fuel elements and unloading the spent fuel elements which has reached its target burnup continuously for reactor operation, the spent fuel elements would be discharged into the spent fuel canister firstly, when the spent fuel storage canister is full of spent fuel, the canister would be sealed through welding method, and then the spent fuel canister would be transferred and stored in the spent fuel storage silo with the ground crane system. The fuel element of the pebble bed high temperature gas cooled reactor is spherical fuel element with graphite matrix, the fuel elements will have friction and collision with the inner wall of the pipeline in transporting process, which will produce graphite dust, the graphite dust should be removed continuously though filtration method, so as not to affect the fuel elements transportation in pipeline. This article focus on the production mechanism and filtering method of the graphite dust in graphite matrix fuel element transporting process in pipeline, to study the graphite dust removal technology, and then we could provide theoretical guidance for the design and operation of the key system and equipment for HTR-PM.

An increasing number of PSA programs concerning research reactors have been launched across the world. As with many other reactors, the CMRR (China Mianyang Research Reactor), a typical pool-type research reactor, regards the control rod shutdown system (CRSS) as its primary shutdown system which enables the reactor subcritical by dropping control rods into the core after a specific initiating event is detected. As a result, the CRSS is an essential ingredient of engineered safety features. It is necessary to enhance the reliability of the CRSS, ensuring the reactor can be successfully shut down when the ATWS — the anticipated transients without scram occurs. Therefore, additional facilities should be designed to cope with the extremely severe circumstance. Accordingly, the purpose of this paper is to evaluate the promotion of the CMRR’s safety degree and the reliability of its CRSS from the PSA’s perspective with an ATWS mitigation system installed. Results indicate that, by introducing the ATWS mitigation system, the failure probability of the CRSS can decrease from 1.52e−05 per demand to 3.35e−06 per demand, while the aggregate CDF (core damage frequency) induced by all IE (initiating event) groups, is able to decrease to a relatively low value 1.17e−05/y from its previous value 3.11e−06/y. It is apparent that the reliability of the CRSS as well as the safety degree of the overall reactor can be enhanced effectively by adding the ATWS mitigation system to the elementary design of the normal CRSS.

Nuclear energy is an important way to solve energy shortage and pollution problems today. Therefore, China is vigorously developing nuclear energy and is facing huge demands for talents. However, nuclear engineering education has been severely hampered, particularly in its experiment aspect because of the Fukushima accident. The original sub-critical nuclear reactor at Tsinghua University (THU) was stopped, forcing students to use computers to conduct relevant nuclear simulation experiments. The emerging of supercomputers and commercial numerical simulation softwares has provided enough hardware and software support for complex calculations required in numerical simulation of nuclear reactors. Thus numerical simulation could be integrated into nuclear engineering education. With its ease of use, quick acquisition and direct visualization of results, numerical simulation can save much time and money. Besides, it is convenient to change simulation conditions which is helpful for academic research and talent development. This paper starts with THU’s nuclear engineering talent training mode, and taking the course “Advanced Nuclear Reactor Thermal Analysis” as an example, discusses the applications of numerical simulation softwares FLUENT and COBRA-TF in this course. Finally, the analysis shows that numerical simulation performs well in conceptual understanding and experimental design. It plays a significant role in nuclear engineering education, which provides an important reference for the new mode of nuclear engineering education.

Chinese Fusion Engineering Testing Reactor (CFETR) is a tokamak-type machine and next device in the roadmap for the realization of fusion energy in China, which aims to bridge the gaps between the International Thermonuclear Experimental Reactor (ITER) and the demonstration reactor (DEMO) [1]. The accident sequence starting from loss of vacuum accident (LOVA) is an important issue concerning the performance of CFETR. During LOVA, air will leak into the vacuum vessel (VV) causing fast pressurization of VV. At the same time, the high speed airflow jet will result in migration and re-suspension of the large quantity of tungsten dust produced and deposited in the lower part of plasma chamber, causing possibilities of radioactive dust leakage into the workshop and environment. In order to conduct a comprehensive analysis of the accident sequence, firstly, the airflow characteristics of LOVA should be studied. In this article, a postulated rupture of different section area is assumed due to a failed component at the equatorial port level. The computational fluid dynamic (CFD) modelling of LOVA was conducted by ANSYS CFX. The results show that the break area has significant influences on the characteristics of the airflow. Two swirling airflows are formed in the upper and lower part of the torus. The airflow characteristics are quite different when the LOVA happens during maintenance or during normal operation. A reverse flow occurs when the LOVA happens during normal operation. Yet can not be observed when LOVA occurs during maintenance. The results are the basis to the further safety study of CFETR such as the re-suspension, migration and explosion of dust.

Steam explosion is one of the consequences of fuel-coolant interactions in a severe accident. Melt jet fragmentation, which is the key phenomenon during steam explosion, has not been clarified sufficiently which prevents the precise prediction of steam explosion. The focus of this paper is on the numerical simulation of the melt jet behavior falling into a coolant pool in order to get a qualitative and quantitative understanding of initial premixing stage of fuel-coolant interaction. The objective of our first phase is the simulation of the fragmentation process and the estimation of the jet breakup length. A commercial CFD code COMSOL is used for the 2D numerical analysis employing the phase field method. The simulation condition is similar to our steam explosion test supported by the ALISA (Access to Large Infrastructure for Severe Accidents) project between European Union and China, and carried out in the KROTOS test facility at CEA, France. The simulation result is in relatively good agreement with the experimental data. Then the effect of the initial jet velocity, the jet diameter and the instability theory are presented. The preliminary data of melt jet fragmentation is helpful to understand the premixing stage of the fuel-coolant interaction.

Immersive Virtual Reality (IVR) systems based on multiple stereoscopic projectors are very popular in many applications, such as training operators for nuclear power plants and surgeons for surgical operations. Due to the increasing number of nuclear reactors in Guangdong province of China, Hong Kong residents are concerned about the nuclear safety and impact on Hong Kong society. There is clearly a strong demand for nuclear literacy education in order to make the public aware of and accept nuclear energy. Thus, City University of Hong Kong has built an IVR system with a 9-meter diameter, 4-meter-height, 235° curved screen for nuclear literacy education. The actual CAD drawings of the Daya Bay nuclear power plant were used to recreate the virtual Daya Bay plant in our IVR system, emphasizing the reactor pressure vessel and steam generators inside the containment building. Visitors can enter the virtual containment building, and experience the actual operation environment in order to understand the basic knowledge of nuclear reactors. At present, the system is not only capable of illustrating the basic knowledge of nuclear reactor physics but also shows the normal and abnormal operations including reactor scram and emergency containment spray. In order to provide visitors with a full understanding of the role of nuclear power in Hong Kong’s fuel mix, a Low Carbon Energy Education Center (LCEEC) was set up in which the IVR system was the main attraction. Other low carbon energy sources are also introduced in LCEEC. The Centre was visited by thousands of visitors since its opening in April 2017. Surveys have been conducted to collect their comments and suggestions. The results showed that the IVR system is very helpful in raising public understanding of nuclear power.

Two kinds of new cladding material for pressurized water reactor were developed by China Nuclear Power Research and Technology Institute (CNPRI) which named CZ1 and CZ2. The cladding tubes of CZ alloys used in the study were fabricated by different final annealing temperature in the range of 450 °C to 600 °C.In order to investigate effect of final annealing temperature on creep property of CZ alloys. The axial creep tests were carried out on specimens of CZ-SRA and CZ-RXA At the temperature of 375 °C with different applied stress levels for 1000 hours. The experimental results showed that the axial creep resistance of CZ-RXA is superior to CZ-SRA at lower stress and is inferior to CZ-SRA at higher stress. The mechanism was discussed.

Breakaway oxidation is a potential fuel failure mechanism during a loss-of-coolant accident (LOCA), especially small-break LOCA. Two kinds of new advanced zirconium alloy, CZ1 and CZ2, were developed by China Nuclear Power Research and Technology Institute of China General Nuclear Power Group. The breakaway oxidation behavior of CZ1 and CZ2 was studied. The outer surface of all samples was examined visually and photographed. After 2730s oxidation in steam, the outer surface of CZ1 alloy sample and CZ2 alloy sample remained lustrous-black. The outer surface of CZ1 sample oxidized in steam at 1000 °C for 4181s was grey, but under the same experimental conditions the outer surface of CZ2 sample was still lustrous black. The hydrogen pickup content of different oxidation time was measured. The samples with grey appearance showed significant hydrogen pickup. The microstructure was observed by optical microscope. Evolution of oxide structure was described, and the mechanism was discussed.

Deep penetration problems exist widely in reactor applications, such as SPRR300 (Swimming Pool Research Reactor 300), a light water moderated, enriched uranium fueled research reactor in China. Deterministic transport theory is intrinsically suitable for deep penetration. But there exist some problems when it’s applied in SPRR-300research reactors. First, the reactor core is complicated for geometry description in deterministic theory codes. Monte Carlo method has advantages in complex geometry modeling. And it uses continuous energy cross sections which are independent with specific reactor types and research objections. But usually it’s difficult to converge well enough to deal with deep penetration problems, even though there are a number of variance reduction techniques. Based on the advantages and disadvantages of Monte Carlo and Deterministic method, we proposed a coupled neutron transport calculation method for deep penetration. It combines advantages of these two methods. Firstly, we use Monte Carlo code to finish fine modeling and do the whole reactor core calculation. Domestically developed Monte Carlo code JMCT is used to do the neutron transport calculation. Then homogenized group constants in each mesh are calculated from JMCT output by a self-developed script. Afterwards, we do the whole reactor calculation with deterministic theory code TORT. It directly uses group constants generated by Monte Carlo code. Finally, we can get the deep penetration calculation results from TORT output. Verification is carried out by comparing the group constants of benchmark problem, and by comparing k eff calculated by this method with continuous energy Monte Carlo method. Benchmark calculation is conducted with OECD/NEA slab benchmark problem. The comparison shows that group constants generated by this study are in good agreement with results from published references. Then above group constants are applied to 3-dimensional discrete ordinates deterministic theory transport code TORT. But k eff calculated by TORT is a little lower than that calculated by Monte Carlo code JMCT. To minimize other influence factors, different Sn/Pn order, and different mesh size in TORT has been tried. Unfortunately the k eff difference between these two methods remains. Even though the k eff results in this benchmark are less than k eff calculated by continuous energy MC method, Benchmark results show that all the group constants generated by this method are in good agreement with existing references. So it can be expected that after further verification and validation, this coupled method can be effectively applied to the deep penetration problem in such kind of research reactors.

The safety of radioactive material transport is regulated by graded approach in China. National requirements on container design, manufacture and shipment are different for the three categories material. Packages of different categories shall be experienced different test to demonstrate their ability to withstand transport normal transport conditions and accident conditions depending on their contents. China Institute for Radiation Protection (CIRP) established free drop test facilities, thermal test facilities, water immersion test facilities and relevant measurement means. CIRP has carried out tests of 18 radioactive material transport packages, and accumulate a lot of experiences about the limit unit analysis, test scheme, test process, monitoring and data treatment. Probabilistic safety assessment method and traffic accident statistics in China are carried out in the recent years by CIRP.

Based on the tri-isotropic (TRISO) fuel technology of high temperature gas cooled reactor, a new fuel concept for improving the accident tolerance of light water reactor (LWRs) named inert matrix dispersion pellet (IMDP) was proposed. Through the silicon carbide matrix and embedded TRISO fuel particles, the safety of the nuclear fuel could be enhanced. Recently, dummy IMDPs were fabricated by China General Nuclear Power Corporation (CGN) and thermal conductivity was tested. According to the tested data, a FEA model using ABAQUS combined its secondary development function was developed and benchmarked. Several influence factors of the effective thermal conductivity (ETC) of the IMDP were studied by the FEA model, such as burn-up, TRISO packing fraction and temperature. The heat transfer behaviors of IMDP and UO2 under typical normal PWR operating condition were also studied.

For the thermal hydraulic and safety analysis of nuclear reactor, a lot of theoretical models and engineering experience are mainly contained in the design software. The degree of self-reliance of software directly reflects the technological level and core competencies. According to the nuclear power plant (NPP) design requirements, self-reliant thermal hydraulic and safety analysis software, such as CORTH, TRANTH, PHYCA and etc., have been developed by Nuclear Power Institute of China (NPIC) with reasonable planning and scientific implementation. In this paper, the development process of the self-reliant software is reviewed, covering requirement analysis, model research, software design, coding, testing, verification and validation. And the main characteristics of self-reliant software were summarized. The successful development of thermal hydraulic and safety analysis software support the export of nuclear power units of China, and enhance the competitiveness.

The simulation of 3D thermal-hydraulic problem for the pool type fast reactors, is one of the necessary and great importance. Most system codes can’t be used to simulate multi-dimensional thermal-hydraulics problems, whereas, the CFD method is suitable to deal with these type of simulation challenges. Based on the CFD method, a neutronics and thermohydraulic coupling code FLUENT/PK for nuclear reactor safety analysis by coupling the commercial CFD code FLUENT with the point kinetics model (PKM) and the pin thermal model (PTM) is developed by University of Science and Technology of China (USTC). The coupled code is verified by comparing with a series of benchmarks on beam interruptions in a lead-bismuth-cooled and MOX-fuelled accelerator-driven system. The variations of transient power, fuel temperature and outlet coolant temperature all agree well with the benchmark results. The validation results show that the code can be used to simulate the transient accidents of critical and sub-critical lead/lead-bismuth cooled reactors. Then this coupling code is used to evaluate the safety performance of MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) at unprotected beam over-power (UBOP) accident, and M2LFR-1000 (Medium-size Modular Lead-cooled Fast Reactor) at the unprotected transient over-power (UTOP) and unprotected loss of flow (ULOF) accident. The transient power, the temperature of coolant and fuel and multi-dimensional flow phenomena in upper plenum and lower plenum are presented and discussed in this paper.

Safety rod and its drive mechanism is one of the shutdown systems in sodium-cooled fast reactor, which must be quickly inserted into the core to achieve emergency shutdown in the event of an accident. Therefore, it is necessary to study the falling process of safety rod. In this paper, the numerical simulation method is used to analyze the falling process of safety rod and its drive mechanism in China Experimental Fast Reactor. According to the flow path of the safety rod and its drive mechanism, the pipe system hydraulic method is used to model the safety rod and its drive mechanism and calculate the hydraulic force of the safety rod and its drive mechanism during the falling process. The relationship between time, displacement, velocity and acceleration is presented. The drop time of safety rod is calculated, which is compared with the experimental results. The factors that affect the drop process are analyzed and a sensitivity analysis is presented.

Nuclear reactor severe accidents can lead to the release of a large amount of radioactive material and cause immense disaster to the environment. Since the Fukushima nuclear accident in Japan, the severe accident research has drawn worldwide attention. Based on the one-dimensional heat conduction model, a DEBRIS-HT program for analyzing the heat transfer characteristics of a debris bed after a severe accident of a sodium-cooled fast reactor was developed. The basic idea of the DEBRIS-HT program is to simplify the complex energy transfer process in the debris bed to a simple one-dimensional heat transfer problem by solving the equivalent thermal conductivity in different situations. In this paper, the DEBRIS-HT program code is prepared by using the existing model and compared with the experimental results. The results show that the DEBRIS-HT program can correctly predict the heat transfer process in the fragment bed. In addition, the heat transfer characteristics analysis program is also used to model the core catcher of the China fast reactor. Firstly, the dryout heat flux when all of molten core dropped on the core catcher was calculated, which was compared with the result of Lipinski’s zero dimensional model, and the error between two values is only 11.2%. Then, the temperature distribution was calculated with the heat power of 15MW.

As one of the generation IV reactors, pool-type Sodium-cooled Fast Reactors (SFRs) is attracting more and more attention. Loss of flow and heat sink accident is one of the most serious accidents for SFRs. Therefore, the decay heat removal capacity after emergency shutdown is of great importance. This paper has developed a one-dimensional code named Decay heat Removal Analysis Code for Sodium-cooled Fast Reactor (DRAC-SFR) to analyze the decay heat removal capacity. In the code, the decay heat removal system contains the primary loop, the intermediate loop and air circuit. The decay heat is removed out step by step with the above three loops. Many studies have been conducted on code verification. The international benchmark analysis of EBRII reactor is applied in the code verification. The calculation is compared with the experimental data and the results of DRAC-SFR agreed well with the experimental data. The comparison with the steady state of China Experimental Fast Reactor (CEFR) shows a good agreement with the design value. The errors of all the compared parameters are within 2%. What’s more, calculation is performed to analyze the characteristics of the decay heat removal capacity for CEFR. Thus, code verification shows that DRAC-SFR is proper to evaluate the decay heat removal capacity for SFRs and has the ability to provide references and technical supports for the design and optimization of the pool-type sodium-cooled fast reactor.

Fire Probabilistic Risk Assessment (PRA) is one of the main methods of fire safety analysis for nuclear power plants (NPPs). At present, the fire PRA under the at-power condition has been widely studied, while the research on the low power and shutdown condition (LPSD) is quite limited. Therefore, in this paper, a second generation NPP on the east coast of China is taken as the research target, and the analysis methods are based on the latest LPSD fire PRA theory in report NUREG/CR-7114. This paper studies the initiating events and ignition frequencies of fire PRA considering the real conditions in LPSD, and established LPSD Fire PRA model, finally obtained the quantitative risk result of the core damage caused by the fire According to the results of this LPSD fire PRA, the fire risk-significant sources and fire risk weakness are found out and the improvement suggestions have been promoted.

Dynamic response research of steam generator loss of coolant accident (SG LOCA) is essential for the reliability and safety consideration. According to the differences of LOCA loading phenomena, two types of LOCA loads affect the SG: rarefaction wave travels through the primary fluid in the U-tubes, and the SG shakes due to reactor coolant loop(RCL) motions transmitted by the primary loop piping, former loading phenomena evaluation is called SG rarefaction analysis while latter is called shaking analysis. This paper place particular emphasis on shaking analysis. At present, the published literatures about LOCA mainly focus on RCL LOCA, reactor LOCA and fuel assembly LOCA, few reports concentrate on shaking dynamic response analysis of SG LOCA. Both Westinghouse and AREVA’s methods according to their research reports are to decouple the SG from the RCL: This method results in low computational efficiency as RCL LOCA and SG LOCA are evaluated separately and the decoupling error is uncertainty, meanwhile, in the vicinity of the nodes where the displacements are imposed, distorted reaction forces are usually found. Through reasonable simplification and equivalence, a detailed nonlinear FEM model of steam generator (SG) of a China 3rd generation nuclear power plant (NPP) is established, this model is then connected with the reactor coolant loop (RCL) to carry out the SG LOCA shaking dynamic response analysis. By calculation, the maximum absolute stresses of SG heat transfer tube bundle and its variation with tube diameter and upper supports reacting forces are obtained. In order to study the effect of SG decoupling from the RCL on shaking dynamic response, a comparative study of decoupling /coupling methods is conducted. Results shows that SG decoupling has a significant impact on the calculation result, the calculation method of coupling is more closer to the real situation and worthy to recommended. Related analytical procedures and calculation results lay the foundation for future SG shaking dynamic analysis and SG design of subsequent power plants.

The high-temperature gas-cooled reactor pebble-module (HTR-PM), which is under constructed in Shidao, Shandong Province of China, adopts the spherical graphite coated components as fuel elements. During the operation of the reactor, the spherical fuel elements will be reloaded into the core periodically through the refueling pipelines, which will cause radioactive irradiation to personnel and facilities in several corridors and rooms. The three-dimensional geometry model was built according to the real structure and the radiation shielding design was verified and optimized by the QAD-CGA program. In the calculation, the average gamma intensities of the spherical fuel elements of 15 different cycle times were adopted as the source term. From the results, it can be concluded that: 1) For the 20-meter corridor, the dose rate will be less than 1E−6mSv/h at 30cm outside the concrete wall of 0.6m thickness by single fuel elements; 2) For the 0-meter corridor, an extra 32.0 cm steel shield should be added to satisfy the personnel dose rate limit of 0.03mSv/h; 3) For semiconductor electrical equipment closed to the pipelines, the required steel shield thickness is 12.0 cm under the accumulative dose limit of 50Gy; 4) For non-radiation resistant cables, the required steel shield thickness is 4.5 cm under the accumulative dose limit of 1000Gy.

Clearance is an effective method to control radioactive waste, but some assessment of types of pollutants may need to be done before clearance in some situation. In fact, there is large volume very low level radioactive materials in nuclear wastes in nuclear power plant, a reasonable method should be built based on both of potential clearance study and clearance method optimization. Some suggestions are presented about the work of radioactive waste disposal for its present condition and development in China.