During a severe accident of a nuclear reactor, radioactive aerosols may be released from degraded nuclear fuels. Pool scrubbing is one of the efficient filters with a high aerosol removal efficiency, in other words a high decontamination factor (DF). Because of its high performance, many pool scrubbing experiments have been performed and several pool scrubbing models have been proposed. In the existing pool scrubbing experiments, an experimental condition of aerosol number concentration was seldom taken into account. It is probably because DF is assumed to be independent of aerosol number concentration, at least, in the concentration where aerosol coagulation is limited. The existing pool scrubbing models also follow this assumption. In order to verify this assumption, we performed a pool scrubbing experiment with different aerosol number concentrations under the same boundary conditions. The test section is a transparent polycarbonate pipe with an inner diameter of 0.2 m. 0.5 μm SiO 2 particles were used as aerosols. As a result, DF was increasing as decreasing the aerosol number concentration. In order to ensure a reliability of this result, three validation tests were performed with meticulous care. According to the results of these validation tests, it was indicated that DF dependence on the aerosol concentration was not because of our experimental system error including measurement instruments but a real phenomenon of the pool scrubbing.

With the advantages of high reliability, high power density and long life, small nuclear power reactor has become one of the most excellent space power options in the space missions. TOPAZ-II is the most mature space nuclear power reactor based on thermionic conversion. In this paper, the thermo-physical and transport properties of NaK-78 and heat transfer correlations for liquid metals are implemented into the RELAP5 code. The modified RELAP5 has already been accessed to analyze the thermal-hydraulic characteristics of the space reactor cooled by NaK-78. A RELAP5 model including the core, TFEs, radiator, coolant loop and volume accumulator is developed. Temperature reactivity feedback, TFE emitter, TFE collector, moderator and the reactivity insertion effects of control drums and safety drums are modeled in the point reactor kinetics equations with six-group delayed neutrons. To V&V the integrated TOPAZ-II system model, the steady state is simulated and analyzed. The steady state calculated results are in good agreement with the designed values. On the basis of V&V, a hypothetical reactivity insertion accident is simulated and analyzed. During the accident, the automatic control system is assumed to be malfunctioned, 0.01$ positive reactivity is introduced for 500s and then control drums start to rotate inward. The maximum temperatures of fuel and emitter are below the melting temperature, respectively. The maximum temperature of coolant is 940K with 160K margin from boiling. With the rotating of control drums, the reactor reaches critical again.

In an extreme situation, as happened in Fukushima nuclear power plant, the failure of multiple safety systems may lead to severe core damage and melt relocation. This may be accompanied by production of large amount of steam which may result in the over-pressurization of the containment. Another associated concern is that the exposed core melt is highly radioactive which if exposed to the atmosphere can ruthlessly deteriorate the quality of environment and living beings. The radioactive materials present in the containment can be in vapor form or aerosol form. The containment of a nuclear power plant is therefore the final shielding to prevent the release of radioactive products to the environment. Therefore, the installation of Filtered Containment Venting system (FCVS) is mandatory in a nuclear reactor which actuates passively to depressurize the containment. Additionally it assists in the retention of radionuclides in the containment. The FCVS consists of venturi scrubbers submerged in a pool of scrubbing liquid along with a demister housed in a scrubber tank. The performance of venturi scrubber is dependent on the interaction of the contaminated air stream from the nuclear reactor with the scrubbing liquid. This represents a multi-field and multi-fluid system. The present analysis investigates this system through a computational framework in which air stream is solved using Eulerian framework while the scrubbing liquid is tracked through Lagrangian framework. The collection efficiency of aerosols is modeled assuming impaction to be the predominant mechanism.

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.

In the accident at the Fukushima Daiichi Nuclear Power Plant, fuel rods and reactor internals were overheated due to the radioactive decay heat and the oxidization of fuel cladding, and relocated to the lower region of the pedestal. To retrieve the fuel debris, location, distribution and characteristics of the fuel debris are quite important. However, this information still has not been clarified. To estimate the distribution of fuel debris, since experiments cannot provide sufficient information, numerical simulation will be a useful tool. There are some numerical codes for calculating conditions inside a reactor at severe accidents, however, those codes include assumptions and/or related to melt relocation, and cannot simulate relocation process directly. In order to simulate the relocation and corium spreading behavior in PCVs without assumptions and scenario, a numerical simulation code that can phenomenologically evaluate the melting phenomena is required. Then, a phenomenologically-based numerical simulation code for predicting the melting core behavior, including solidification and relocation based on the computational fluid dynamics, JUPITER, has been developed in JAEA. In this paper, by using JUPITER, we simulated the corium spreading process inside a pedestal. As a result, we confirmed that JUPITER has potential to estimate the debris distribution inside the pedestal region.

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.

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.

The ENEN+ project proposes cost-effective actions to attract, develop and retain new talents in nuclear professions. This is a contribution of the ENEN Association, supported by the European Commission, to the common strategic goal of all nuclear stakeholders: to preserve, maintain and further develop the valuable nuclear knowledge for todays and future generations. The ENEN+ project focuses on learners and careers in nuclear reactor engineering and safety, waste management and geological disposal, radiation protection and medical applications. This paper outlines the most important expected contributions of the project and the approaches designed to deliver them.

A square array of cylinders subjected to axial flow is commonly encountered in nuclear reactors and other heat exchangers. Large-scale vortices have been observed in the gaps between the cylinders, both experimentally and numerically. These periodic flow instabilities occur in tightly-spaced cylinder arrays and originate from the velocity difference between the gap and the subchannel regions. The pressure fluctuations caused by the coherent vortex structures are possibly a source of fretting and fatigue in the aforementioned applications. In order to quantify and comprehend this phenomenon, Large-Eddy Simulations are performed on an incompressible, Newtonian fluid flowing adiabatically through a numerical domain containing a single rigid cylinder with periodic boundary conditions, representative for a cylinder in an infinite square array. Subsequently, the temporal frequency spectrum of the wall pressure profile is calculated. The spatial autocorrelation function of this Fourier spectrum, the so-called Cross Spectral Density function, contains information regarding the amplitude and convection speed of the pressure fluctuations. It is shown that the flow instability is strongest for a pitch-over-diameter ratio of 1.03. Also, the simulations indicate that the convection speed is monotonously increasing with the pitch-over-diameter ratio. An updated model for this convection speed is proposed. Finally, it is shown that the single-cylinder approximation has its limitations, but provides valuable information with minimal computational cost.

Computational Fluid Dynamics (CFD) methods have been developed into effective fluid simulation means to be used on the hydraulic design in the field of nuclear reactor. However, it is difficult to generate suitable mesh and select turbulence model for simulation because of the complex geometry structure and flow behavior in the pressure vessel. Based on CFX software, a hydraulic computation model of typical pressurized water reactor is established and the flow distribution at core inlet is analyzed. The simplified geometry model consists cold legs, downcomer, lower plenum, secondary support component, core support plate, and lower core plate. The computation model is divided into three parts for mesh generation, including the part of inlet and downcomer, the part of lower plenum and core inlet section, and the part of core. In order to reach the independence of grid several methods of mesh generation which contains different mesh density at local key parties are investigated to screen out the suitable mesh scheme. The k-ε, k-ω, and SST k-ω turbulence model are respectively used for simulation and the sensitivity of turbulence model at different locations of flow field is analyzed. The results show that the mass flow rate of the near wall flow field, computed by using k-ω turbulence model, is consistent with SST k-ω model, while the mass flow rate of central flow field computed by using k-ε turbulence agrees with SST k-ω model. The result computed by using k-ε turbulence model shows relatively uniform flow distribution at core inlet, which is more consistent with the measured data with the average difference of 3.1%. By using the k-ε model, the probability distribution of the difference between the calculated results and the experimental values follows the law of Normal distribution. The final coolant flow distribution at each orifice is evaluated, and the maximum normalized flow flux is found at center orifice while the flow rate at the edge of core is relatively lower.

Global-local self-shielding calculation scheme is a new high-fidelity resonance calculation model proposed by NECP laboratory of Xi'an Jiaotong University. Neutron Current Method (NCM) is utilized for resonance calculation in the global aspect to obtain Dancoff factors. Then each fuel pin is transformed into individual 1D cylindrical problems by conserving Dancoff factors. The Pseudo-Resonant-Nuclide Subgroup Method (PRNSM) is used to conduct resonance calculation in the local aspect for each 1D cylindrical pin. Global-local self-shielding calculation scheme has been successfully implemented in high-fidelity numerical nuclear reactor physics code NECP-X. Verification results of global-local self-shielding calculation scheme showed good accuracy for UO2 fuels. The maximum relative error of microscopic absorption cross sections (XSs) for 238 U in resonance range was 1.5% compared with MCNP5 [1]. AIC control rods serve as strong absorbers in reactor. Strong self-shielding phenomenon occurs when AIC control rods are inserted. Analysis was performed to determine the effects of AIC control rods on the accuracy of global-local self-shielding calculation scheme and the sources of error. Evaluation results showed that the main part of error was introduced by NCM and radius searching. The relative errors were larger than 10% in several resonance groups. Therefore, a supercell model is proposed to couple with global-local self-shielding calculation scheme to treat resonance calculation for AIC control rods in this paper. Numerical results show that this model improves the accuracy of the global-local self-shielding calculation scheme. The relative errors of microscopic absorption XSs for AIC in most resonance groups were decreased to less than 2%.

Monte Carlo (MC) burnup calculation method, implemented through coupling neutron transport and point depletion solvers, is widely used in design and analysis of nuclear reactor. Burnup calculation is generally solved by dividing reactor lifetime into steps and modeling geometry into numbers of burnup areas where neutron flux and one group effective cross sections are treated as constant during each burnup step. Such constant approximation for neutron flux and effective cross section will lead to obvious error unless using fairly short step. To yield accuracy and efficiency improvement, coupling schemes have been researched in series of MC codes. In this study, four coupling schemes, beginning of step approximation, predictor-corrector methods by correcting nuclide density and flux-cross section as well as high order predictor-corrector with sub-step method were researched and implemented in RMC. Verification and comparison were performed by adopting assembly problem from VERA international benchmark. Results illustrate that high order coupled with sub-step method is with notable accuracy compared to beginning of step approximation and traditional predictor-corrector, especially for calculation in which step length is fairly long.

In this paper the application of the virtual element method (VEM) to the multigroup, neutron diffusion equation will be presented. The VEM is a recently developed Bubnov-Galerkin spatial discretisation method based largely on the mimetic finite difference method (MFD) that preserves the properties of the underlying vector operators. It can discretise elliptic partial differential equations (PDEs) on arbitrary polygonal/polyhedral meshes with arbitrary order and regularity. Deterministic, geometry conforming methods, used to solve multigroup, neutron diffusion, nuclear reactor physics problems have historically used the finite element method (FEM). However, FEM requires high-quality meshes with few highly distorted elements (elements with a poor aspect ratio) or it may may experience convergence problems. The process of creating high quality meshes, even with automated mesh generation algorithms, such as the advancing front and Delaunay methods, is often very time consuming. For these reasons VEM is being studied in this paper as a possible alternative to FEM in the numerical solution of neutron diffusion problems in nuclear reactor physics. A C5G7 UOX pincell problem is presented to demonstrate the application of VEM to mutligroup diffusion problem. The method of manufactured solutions (MMS) is used to determine the order of convergence.

Until recently, interior penalty methods have been applied to elliptic operators using an approach based on the mass matrix of finite elements that possess a constant Jacobian. In the case of isogeometric analysis, such an approach would not always guarantee coercivity of the bilinear form and thus numerical stability of the solution. In this paper, optimal interior penalty parameters [1] are described and applied to the symmetric interior penalty scheme of the discontinuous Galerkin isogeometric spatial discretisation of the neutron diffusion equation. The numerical accuracy of the proposed method is compared against a standard continuous Bubnov-Galerkin isogeometric spatial discretisation of the neutron diffusion equation. Numerical consistency and order of error-convergence is verified by means of the method of manufactured solutions. Numerical results are also presented for a two-dimensional pin-cell test case, based upon the OECD/NEA C5G7 quarter core MOX fuel assembly benchmark for nuclear reactor physics parameters of interest.

Radiation-induced damage and degradation to shielding and structure materials in nuclear reactors is one of the key limiting factors that affect the safety considerations and the lifetime. Neutron radiation damages materials mainly by exciting a number of Primary Knock-on Atom (PKAs). PKAs induce displacement cascades, causing microstructure changes and mechanical degradations in materials. Computer simulations are used to model this complex process. Knowing the PKA spectrum accurately is important because PKA spectrum and the input of computer simulation are coupled. In this work, we aim to obtain the PKA spectrum in Zr-based alloys with relatively low Zr concentration by using the Geant4 software. Some new functions were added by reprogramming in Geant4. We found that the energy spectra of PKA in Zr 2 Cu and Zr 2 Ni are mainly caused by Zr atom, and the shape of the average energy spectra are similar with pure Zr. The number of PKA distributions and the energy deposition in these two Zr-based metals are similar with pure Zr but different than those in pure Ni. These finding indicate that the metallic elements of Cu and Ni have small impact on PKA spectrum in Zr-based alloys, which illustrate that simplified simulation models are feasible when using computer for simulating. Moreover, it has great significant for the calculation of irradiation damage and the computer simulation for the process of collision cascade after one PKA is formed.

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.

ThO 2 has been considered as a possible replacement for UO 2 fuel for future generation of nuclear reactors, and thorium-based mixed oxide (Th-MOX) fuel performance in a light water reactor was investigated due to better neutronics properties and proliferation resistance compared to conventional UO 2 fuel. In this study, the thermal, mechanical properties of Th 0.923 U 0.077 O 2 and Th 0.923 Pu 0.077 O 2 fuel were reviewed with updated properties and compared with UO 2 fuel, and the corresponding fuel performance in a light water reactor under normal operation conditions were also analyzed and compared by using CAMPUS code. The Th 0.923 U 0.077 O 2 fuel were found to decrease the fuel centerline temperature, while Th 0.923 Pu 0.077 O 2 fuel was found to have a bit higher fuel centerline temperature than UO 2 fuel at the beginning of fuel burnup, and then much lower fuel centerline than UO 2 fuel at high fuel burnup. The Th 0.923 U 0.077 O 2 fuel was found to have lowest fuel centerline temperature, fission gas release and plenum pressure. While the Th 0.923 Pu 0.077 O 2 fuel was found to have earliest gap closure time with much less fission gas release and much lower plenum pressure compared to UO 2 fuel. So the fuel performance could be expected to be improved by applying Th 0.923 U 0.077 O 2 and Th 0.923 Pu 0.077 O 2 fuel.

Continuous improvement in Nuclear Industry Safety Standards and reactor designer’s and operator’s commercial goals lead to an increasing demand of fast running and highly accurate methodologies oriented to improve the prediction capabilities of main reactor’s parameters under steady state and transient situations. An increasing effort has been observed during past years to develop high accurate multi-physics approach for nuclear reactor analysis, based both on the availability of advanced codes and the constant increase of computational resources with massive parallel architectures. As a result, several improvements have been observed in the implementation of coupled full 3-D Monte Carlo (MC) neutronic models for nuclear reactor cores, including not only the coupling to thermal-hydraulics but also fuel behavior codes. This approach has proved at concept level to be able to develop high accurate models that would allow to predict important safety and performance parameters of nuclear reactors with less conservativism. Under this framework, the European Research and innovation project McSAFE is a coordinated effort started in September 2017 with the objective of moving MC stand-alone and coupled solution methodologies to become valuable tools for core design, safety analysis and industry like applications for LWRs of gen II and III. In this work the Serpent 2 code, a high performance MC code developed by VTT, is used by the aim of performing the preliminary screening of capabilities, performance and limitations of such challenging objective. As a result, simplified analysis are developed to identify full 3-D modeling computational requirements for typical LWR configurations, including burnup aspects. Potential bottlenecks and limitations are presented and discussed, providing foreseen alternatives and solutions for further code improvements.

In order to ensure the safety of fuel rods in nuclear reactor, it is necessary to consider the condition of the power ramp during reactor operation, which may cause breakage risk of fuel rod at the time of pellet-cladding interaction (PCI) appearing. To analyze this phenomenon and reduce the risk, a performance analysis model for fuel rod is developed to carry out the steady state and transient simulation using commercial software COSMOL. The full model has three main computational models, which are heat transfer model, mechanical model and internal pressure model. The calculation results show that the stress and strain of NHR200-II fuel rods has enough margins during the power ramp process, which indicates that the risks of fuel rod damage are very low.

The efficient solution of the neutron diffusion equation for large scale whole core calculations is of paramount importance; especially if the detailed pin-level power distribution and reaction rates are required. For heterogeneous whole core calculations finite element based techniques have been one approach to modelling the detailed pin geometry in whole core calculations. A new approach, pioneered in the last few years, is isogeometric analysis (IGA) methods which enable the exact geometry of fuel pins to be modelled. In order to efficiently solve elliptic partial differential equations (PDEs), such as the neutron diffusion equation, typically multi-grid or multi-level iterative solution techniques are used such as the algebraic multi-grid method. However, using IGA methods it is possible to develop true geometric multi-grid techniques which are potentially much more efficient than the standard algebraic multi-grid methods. In this paper we explore the use of IGA methods to develop a scalable, multi-level, iterative algorithm which is then used to solve the neutron diffusion equation over several geometries. This multilevel solution algorithm utilises a single patch multi-grid framework suggested by Hofreither and Takacs that takes advantage of the tensor product construction to provide scalability with respect to spatial, and polynomial refinement. Furthermore, a two-level balancing Neumann-Neumann solver is used to extend the solver to multiple patches in a scalable way. It is seen that the number of iterations required depends on the mapping between the unit square and the physical geometry, as well as the material coefficients.