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.

Molten salt reactors are a very promising option for the development of highly innovative solutions for the nuclear energy production of the future. The techniques used to model thermal hydraulics of a molten salt fast reactor when frozen salt wall technology is applied to the core vessel wall are presented here for 2D numerical models of a hyperboloid reactor core region with a heat exchanger was applied in Code_Saturne. A 3D simulation of the fluid flow and heat transfer with 16 recirculation loops containing the heat exchangers is also presented. It was found that there is strong cooling in separated flow regions in the external heat exchanger, which freezes where the porous model is applied.

The International Nuclear Management Academy (INMA) is an International Atomic Energy Agency (IAEA) framework to support the establishment and sustainability of Master’s level Nuclear Technology Management educational programmes and the development of nuclear technology management professionals. The INMA framework describes a broad range of competencies across four Aspect Groups of External Environment, Technology, Management and Leadership, that have been identified as the basis for the successful management of nuclear projects. By following the INMA framework these competencies can be achieved by nuclear technology subject matter experts to support their career path into managerial roles or by experienced managers moving into the nuclear sector. The IAEA in conjunction with worldwide universities with nuclear education programmes have developed an endorsement process to recognise which university Master’s programmes adhere to the INMA framework and can therefore produce graduates with the required competencies. It is also recognised though that the implementation of these competencies can only be fully achieved through on-the-job training or experiential learning. A combination of education and experience is therefore required to be recognised as a nuclear technology management professional. To date two universities, The University of Manchester and the Moscow Engineering Physics Institute, have received INMA endorsement for their Master’s programmes in Nuclear Technology Management. The University of Manchester programme is part-time while the MEPhI programme is a two-year full-time programme. Several other universities — North West University and University of the Witwatersrand (both South Africa), Texas A&M University and the University of Tokyo having been assessed for endorsement, and many others developing nuclear technology management programmes are entering the process. The IAEA organise an INMA Annual Meeting where universities can meet to express interest in the programme, learn more about what is required for the programme and endorsement, and exchange best practices. The International Nuclear Management Academy is therefore making significant contributions to improving nuclear technology management competencies leading to improved managerial decision making with the associated benefits to the global nuclear industry.

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.

The European Nuclear Education Network (ENEN) was established in 2003 through an EU Fifth Framework Programme (FP) project, as a legal nonprofit-making body. Its main objective is the preservation and further development of expertise in the nuclear fields by higher education and training. This objective is realized through the cooperation between EU universities involved in education and research in nuclear disciplines, nuclear research centers and the nuclear industry. As of March 2018, ENEN has 66 members in 18 EU countries and has concluded Memoranda of Understanding (MoU) with partners beyond Europe for further cooperation, including organizations in, Russian Federation, Ukraine, Canada and Japan. ENEN also has good collaboration with national networks and international organizations such as the Belgian Nuclear Education Network (BNEN) and the International Atomic Energy Agency (IAEA). The main activities developed, and results achieved, within the first 15 years of the ENEN Association are presented and discussed. These include, for example, the launch of the European Master of Science in Nuclear Engineering (EMSNE), the annual ENEN Ph.D. competition and the portfolio of more than 10 EURATOM projects dealing with nuclear education, training and knowledge management through development of teaching methods and materials, courses, and exchange of students and teachers within EU and beyond. Those projects were all supported by the European Commission with the ENEN Association acting as the coordinator or as a partner.

The United Arab Emirates (UAE) is a developing affluent nation. The leaders of the UAE announced the pursuit of peaceful nuclear power in 2008 and by the end of the following year established its Nuclear Energy Program Implementing Organization (the Emirates Nuclear Energy Corporation (ENEC)), Federal Authority for Nuclear Regulation (FANR), and ordered four APR-1400 pressurized water reactors from the Korean Electric Power Company (KEPCO). Nuclear Engineering programs were initiated soon afterwards at Khalifa University for graduate students and the University of Sharjah for undergraduate students. The technical workforce including nuclear power plant local operators and chemistry and radiation protection personnel was established by ENEC and the Institute of Applied Technology as an inaugural program of Abu Dhabi Polytechnic (AD Poly) in 2011. This paper describes the development of the dual education and training program at AD Poly, the experience of the initial cohorts who conducted their training at the APR-1400 units at the Shin Kori Nuclear Power Plant in Korea, and the current program between the AD Poly Abu Dhabi campus and the new Barakah Nuclear Power Plant based on lessons learned from the earlier years.

In nuclear reactors that use plate-type fuel, the fuel plates are thermally managed with coolant flowing through channels between the plates. Depending on the flow rates and sizes of the fluid channels, the hydraulic forces exerted on a plate can be quite large. Currently, there is a worldwide effort to convert research reactors that use highly enriched uranium (HEU) fuel, some of which are plate-type, to low-enriched uranium (LEU). Because of the proposed changes to the fuel structure and thickness, a need exists to characterize the potential for flow-induced deflection of the LEU fuel plates. In this study, as an initial step, calculations of Fluid-Structure Interaction (FSI) for a flat aluminum plate separating two parallel rectangular channels are performed using the commercial code STAR-CCM+ and the integrated multi-physics code SHARP, developed under the Nuclear Energy Advanced Modeling and Simulation program. SHARP contains the high-fidelity single physics packages Diablo and Nek5000, both highly scalable and extensively validated. In this work, verification studies are performed to assess the results from both STAR-CCM+ and SHARP. The predicted deflections of the plate agree well with each other as well as exhibiting good agreement with simulations performed by the University of Missouri utilizing STAR-CCM+ coupled with the commercial structural mechanics code ABAQUS. The study provides a solid basis for FSI modeling capability for plate-type fuel element with SHARP.

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.

In this paper, preliminary neutron physical properties of ceramic fast reactor (CFR) are simulated and analyzed. The CFR core consists of ceramic materials, including nuclear fuels, coolants, structural materials, reflective and absorption materials. These ceramics improve inherent safety levels substantially, increase breeding performance, and enhance the power-generation efficiency. The CFR has the potential to operate and breed more than 30 years. The performance of the CFR was simulated focusing on neutron-related effects. The parameters discussed contain fast neutron energy spectrum, the ideal effective multiplication-factor, nuclides mass changes, breeding performance, operation mode, etc. Furthermore, the strengths of the proposed reactor system are discussed. In the future nuclear energy system, CFR may be one of the existing alternative novel reactor type.

Heat pipe-segmented thermoelectric module converters space reactor power system (HP-SMTCs SRPS) is a promising candidate for space nuclear power system. An examination was taken to discuss the criticality safety of HP-SMTCs reactor core in several accident conditions. In the original nuclear design, the case that reactor core is submerged in wet sand while the voids inside the core are filled with sea water and the BeO reflector are dismantled is defined as the worst status with the highest risk of supercritical. However, recent Monte Carlo transport calculation result shows that reactivity in the case that the core is submerged in water while the voids inside are empty is even higher than those cases with the voids full of sea water, which means that the reactor may encounter high risk of supercritical when some particular accidents occur. Detailed analysis about the neutron energy spectrum and absorption reaction rate is made in order to find out the potential reason of these unexpected results. According to the discussion about the criticality safety issues in some accidents, further evaluations may be necessary for the neutronics design of HP-SMTCs space reactors.

Nuclear power units need to operate conditioned the lowest risk possible. Safety analysis must use paired models, combining probabilistic and deterministic methods. In this study, FRAPCON and FRAPTRAN codes were used to simulate an idealized test based on IFA-650 series, carried out within Halden program. Nuclear systems work to depend on uncertainty values that must be quantified and propagated. The sources of uncertainties can be divided among physical models, boundary conditions, and mechanical tolerances. Eight physical models that can be configured, such as thermal conductibility, and fission gas release. Mechanical tolerances introduced by fuel fabrication are deviations that must propagate throughout of the system. To measure the effects produced by uncertainties were used correlation coefficients between entry and exit. Uncertainties contained on input values are spread to measure the impact created on safety limits. The method adopted used 96 samples to achieve the 95% of probability and 95% of confidence level.

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.

According to the development of the concept of “zero failure” or “zero fuel element defect”, accepted in 2011, which consists in reducing the number of fuel elements that are depressurized in the process of operation to the reached level in the leading countries in nuclear energy (10 −6 –10 −5 defective fuel rods) and avoidance of fuel assemblies with non-hermetic cladding of fuel rods for further operation, including defects with a “gas leak” type, new promising fuels are being developed and introduced, including methods for justifying their safety. Thus, to ensure reliability and safety of new fuel types, it is necessary to provide procedures for monitoring current performance characteristics at all stages of the life cycle of fuel rods. In this paper, experience is given on the development and implementation of instrumentation and methods for monitoring of fuel rods with advanced types of nuclear fuel for VVER reactors that ensure the reliability, safety and competitiveness of technologies associated with the use of advanced fuel rod types, and the implementation of associated components, systems and equipment for monitoring and diagnostics. The features of the applied techniques are presented, and the new system of requirements for the implemented equipment created on their basis. This research continues, and the analysis of intermediate experimental data is carried out in this article.

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.

When a severe accident of a nuclear reactor occurs, decommissioning work becomes important task. In the decommissioning work of a boiling water reactor after the severe accident, estimation of the sedimentation place of the molten debris is important. However, the technique to estimate exactly the sedimentation place has not been enough developed. Therefore, the detailed and phenomenological numerical simulation code named “JUPITER” is under development for predicting the molten core behavior in JAEA (Japan Atomic Energy Agency). The comparison between experimental and numerical results is necessary to clarify the validity of the numerical analysis code. The study provides the experimental data to examine the numerical simulation code. As a basic study to examine the numerical simulation code, a liquid film flowing in a modeling flow channel was studied by using water. The flow was visualized, and the flow data were obtained by image processing.

The outcomes of successive droplets impacting onto solid surface of the steam separator in a nuclear power system’s steam generator has a strong effect on the separating efficiency. Due to amounts of influencing factors, experimental research is an important method to study the phenomena of droplet-wall collision. However, because it is hard to generator continuous droplets with controllable uniform size and frequency, experimental studies about successive droplets impacting on solid surface are relative limited. In this study, a novel drop-on-demand (DOD) droplet generator is designed and fabricated based on piezoelectric ceramics, in which successive droplets with a uniform diameter can be generated. Firstly, the structure design of the DOD droplet generator, the setup of the control system and working principle are described in detail in this paper. Then the droplet generating performance of the device under different signal frequency f s , signal amplitude U , duty ratio D r , and nozzle diameter D n are investigated experimentally using a high-speed camera at 4000 fps. Finally, the influence of the signal frequency f s , voltage U , duty ratio D r and nozzle diameter D n on the diameter of droplet D d is discussed. A test of successive droplets generated by the device impacting on an aluminum plate is conducted.

For nuclear power system, the supercritical CO2-based Brayton cycle is very promising for its potentially higher efficiency and compactness compared to steam-based Rankine cycle. Compressor is the critical component in the supercritical CO 2 -based cycle, which typically operates at an inlet fluid state close to the fluid critical point for optimal cycle efficiency. As the fluid parameters vary significantly near the critical point, the compressor is more vulnerable to flow instabilities and care must be taken in designing the compressor. The supercritical CO2 radial compressor features a compact design and the unsteady interactions between the impeller and the vaned diffuser are typically strong. A comprehensive understanding of the unsteady flow effects in the compressor is very helpful in guiding the aerodynamic design. However, little work has been performed on the flow analysis of the unsteady impeller-diffuser interactions in the supercritical CO2 radial compressor. In this work, the unsteady flow simulation of a supercritical CO2 radial compressor stage is carried out. Strong flow unsteadiness is observed and the isentropic efficiency shows a variation of over 20% within one revolution.

Korea Atomic Energy Research Institute (KAERI) has designed and constructed a test facility for reactor coolant pumps (RCPs). In RCP test facility (RCPTF), it is possible to perform a type test for a newly-developed RCP as well as a production test for the Advanced Power Reactor 1400 (APR1400) RCP before its installation in nuclear power plants. The thermal hydraulic and electric capability of the RCPTF covers up to 18.5 MPa, 343 °C, 11.7 m 3 /s, and 14.0 MW for the design pressure, temperature, flow rate, and maximum electric power, respectively. In 2013, a commissioning test was performed to verify its designed capability, followed by several modifications in the RCPTF including signal processing and control logic to enhance the verification and evaluation capability of the RCP performance. During the commission test period, the technical activity for RCP performance verification test and optimization of flow stability have also been performed. The developed techniques can be divided into four subjects; 1) Development of flow stabilization technique for high flow condition test facility, 2) Improvement on the accuracy of performance verification factor and development of evaluation method, 3) Evaluation of test facility vibration characteristics and optimization of measurement system, and 4) RCP coast-down data production. KAERI has completed a full set of technique developments, a prerequisite for the RCP performance test.

The flow rate can fluctuate in offshore nuclear power systems which are exposed to wind and waves, as well as in loops where flow instabilities occur, resulting in different thermal-hydraulic characteristics compared with that under steady flow. Among the thermal-hydraulic characteristics, onset of nucleate boiling (ONB) model determines whether the fluid is boiling, and boiling heat transfer is crucial to equipment performance and safety, both being key issues in subcooled flow boiling. Therefore, an experimental study was conducted to investigate how an imposed periodic flow oscillation affects the boiling inception and heat transfer of subcooled flow boiling of water in a vertical tube. The experiments were conducted under atmospheric pressure with the average flow rate ranging from 96kg/m2s to 287kg/m2s and heat flux ranging from 10kW/m2 to 197kW/m2. The relative pulsatile amplitude range is 0.1–0.3 and pulsatile period range is 10s-30s. Photographic images and thermal parameters such as temperatures and flow rate were recorded. The lack of nucleation site on the heated surface of the test section results in high wall superheat at ONB. The effects of pulsatile amplitude and period on superheat at boiling onset and average heat transfer were analyzed. The results show that the superheat at boiling inception is decreased when the average heat flux is lower than the heat flux at boiling inception of the corresponding steady flow, and the superheat at boiling onset is increased when the average heat flux is higher than the heat flux at boiling onset of the corresponding steady flow. The above effect of flow rate pulsation on superheat increases with increasing amplitude and decreasing period, and the mechanism can be explained by boiling nucleation theory. The lack of large active nucleation site also affects the boiling heat transfer. By comparing the contribution of nucleate boiling to heat transfer with the widely used Cooper’s pool boiling correlation, the subcooled flow boiling was found suppressed by convection. The average heat transfer of both the intermittent flow boiling and the single phase flow is influenced by flow oscillation.