SiC has become a candidate cladding material of Accident Tolerant Fuels (ATF) due to its excellent irradiation stability and corrosion resistance. However, because SiC is a ceramic material with low toughness, brittle failure is a significant concern. In order to improve the toughness, SiC fiber is required to manufacture multi-layer SiC composites. But the current performance model or analysis tool is not available for SiC composites cladding due to its obviously difference with Zr alloy cladding. On one side, Finite element method was used in this paper to analyze the performance of SiC composites cladding under operation conditions which include normal, transient conditions and LOCA conditions; on the other side, this paper gives the performance of the SiC composites with two layers under multiple operating conditions. The result showed that the temperature was stable and the maximum hoop stress was reached at about 70d under normal condition. The power ramp can increase the cladding temperature and has visible influence on the stress distribution. The hoop stress of the cladding reversed under LOCA condition. The tensile hoop stress on the outer surface significantly increased, which caused the obvious increase of failure probability of monolithic SiC, and the failure probability of SiC layer is significantly increased. The conclusion of the analysis has guiding significance for the theoretical design of SiC composites.

Present study is focused on the computational analysis of melting of PCM inside the spherical capsule. Both unconstrained and constrained melting is analyzed for the constant PCM volume and similar initial and boundary conditions. RT27 is chosen as the PCM for this study. Air is considered at the top of PCM inside the spherical capsule. Results are validated with the existing experimental and computational results and found to be in good agreement. Results obtained from present study are compared for the melting fraction, pattern and time. Composite diagrams are presented for the streamline and temperature contours.

UO2-BeO composite fuel may enable Light Water Reactors (LWRs) to have better safety due to its higher thermal conductivity. Much work have been done on the analysis of UO2-BeO fuel performance during LWRs steady state and Loss of Coolant Accident (LOCA) conditions using hypothetical thermal properties and behaviors models, leading to much uncertainty of the results. In this paper, firstly, fuel swelling and densification models for UO2-BeO fuel were developed based on Halden experiment data. Secondly, UO2-BeO fuel thermal properties and behaviors models have been coded in FRAPCON4.0 and FRAPTRAN2.0 after an evaluation of their applicability to UO2-BeO performance simulation. Then, both UO2-BeO composite fuel and traditional UO2 fuel performance during normal conditions and RIA were done in this paper by modified version of FRAPCON4.0 and FRAPTRAN2.0. Finally, comparisons between UO2-BeO and UO2 performance were conducted. The results shows that the peaking temperature of fuel can be reduced about 200K and 150K during normal conditions and RIA by adopting UO2-BeO, respectively. At the same time, the onset of pellet-cladding mechanic interaction (PCMI) can be delayed about 100days during normal conditions and the weakened PCMI effect can be expected during reactivity insertion accidents (RIA) due to the lower thermal expansion coefficient and temperature distribution for UO2-BeO composite fuel. Also, enthalpy stored in UO2-BeO fuel is reduced about 1/5 compared with that of UO2. However, fission gas release ration of UO2-BeO was a bit larger than that of UO2 due to its higher average burnup. And, further experiments stilled required to gain data for UO2-BeO during high burnup, like possibly reduced thermal conductivity and fission gas release threshold.

UO 2 -BeO is one of the most promising accident tolerant nuclear fuels due to its excellent thermal conductivity compared to pure UO 2 fuel. Two different UO 2 -BeO fabrication methods have demonstrated the capability to fabricate enhanced thermal conductivity UO 2 -BeO composite fuels and improve fuel performance. In one method, BeO is continuously distributed around UO 2 grains while BeO is dispersed in the UO 2 phase in the other method. In the former type, BeO is considered as matrix and UO 2 as dispersed particle. However, in the later type, BeO is considered as dispersed particle and UO 2 as matrix. To calculate the thermal conductivity of UO 2 -BeO composite, Hasselman-Johnson model has been applied, which shows good agreement with experimental data. In this model, it includes the influence of thermal conductivity of matrix and particle, volume fraction of particle, radius of particle and the interfacial thermal conductivity between matrix and particle. To balance the improvement of thermal conductivity and enrichment of UO 2 , a UO 2 -BeO composite fuel with 10% volume fraction of BeO has been chosen. Besides, the grain size of matrix should be noticeable smaller than particle to relax thermal stress which may cause micro-cracks and destruction, leading to the grain sizes of UO 2 in two types being distinctive, and resulting in very big effect on fission gas release. In the paper the thermal conductivity has been intensively studied as well as fuel performance in two different types of UO 2 -BeO fuels, and the two fabrication methods have also been compared to assess their applications in commercial reactors.

The radial porosity generally have a higher value at the container wall than that in the core part. Consequently, the fluid flows are mal-distributed in packed bed with significant bypass flow at the wall, which lowers the convective heat transfer performance inside the packed bed. To overcome this drawback of packed bed, we developed an effective way to construct the radial layered composite packed bed, which can easily realize placing small particle at the near wall region and large spheres in the core region. Therefore, smaller pores forms close to the container wall and larger channels presented in the core part. This could result in a much homogenous radial porosity distribution, which is benefit to restrain the bypass flow near the wall. In present paper, the packing procedure is simulated by the discrete element method (DEM). Radial layered composite packed bed and traditional packed bed with uniform spheres are compared on radial porosity distribution. By altering the size of spheres ratios in the near wall region different radial layered packings are also generated and compared. Then, the geometries of these packed bed are imported into the computational fluid dynamics (CFD) simulation. The fluid flow inside the packed bed is investigated. It finds that the radial layered packed bed has a lower pressure drop with the ordered packing structure. And a much homogenous fluid flow distribution is obtained than the traditional, which is benefit for the heat removal inside the packed bed. This would be useful for the optimum design of packed bed in industry applications.

No experimental investigation has been carried out to understand the progression of channel disassembly in IPHWR. Hence, it is planned to study the deformations aspects through experiments with a scaled-down (1:3) fuel channel of 220MWe reactor. Depending upon the pressure inside the channel and transient past, the accident initiation and disassembly can be can be classified into two class. In the first case, high pressure inside channel leads PT ballooned and fully contacted CT before channel deformed as one composite unit. While in second case, low inside pressure, first leads PT to contact CT and then coupled deformation of the channel progresses. Latter is part of our current experimental investigation. This paper is focused on discussion about various aspects considered while developing the experimental facility. In this case, experiments are grouped under two broad categories, one is the study under inert environment to check channel deformation due to material flow at higher temperature and other one is under oxidation environment, which would be similar to actual condition exist in normal reactor core under accident scenario. Initially experiments will be carried out on single channels, later it will be extended up to three channels arranged one above other. A simulation study is also carried out to check the deformation profile under inert conditions. Simulation is done using coupled field element technology of commercial software ANSYS v17.2, to get a preliminary insight into structural and thermal coupled response on channel deformation.

Steel Concrete Composite Wall (SC-Wall) is a unique composite structural component designed and used for shield building wall in new nuclear power plants construction. The composite function (action) of steel-plate and concrete is through the use of embedded studs, which are designed in an appropriate pattern to secure the connections between concrete and surface steel-plates. Thus the behavior of studs (primarily shear behavior) determines the composite state and its functions for serving as an integrated one-piece (or monolithic) section. For elastic state (linear) and ultimate state design, ACI 318 / 349 and recent published design specifications provide formula / equations for shear design-evaluations between plate stud and concrete; but for Beyond Design Basis (BDB) loading conditions, since the section behaves in elasto-plastic (or plastic) state, the constitutive relations among plate-stud-concrete have not yet established in current prevailing codes / or standards. Considering this situation, this paper is contributed to investigate the section behaviors of SC wall subjected to above BDB seismic loadings, provide recommendations for constitutive relations to reveal the actual strain-stress conditions and composite states, and further provide criteria suitable for section design and evaluations under such a circumstance. This research and investigation are based on an analytical methodology and physical model.

In order to maintain the U.S. domestic nuclear capability, its scientific technical leadership, and to keep our options open for closing the nuclear fuel cycle, the Department of Energy, Office of Nuclear Energy (DOE-NE) invests in various R&D programs to identify and resolve technical challenges related to the sustainability of the nuclear fuel cycle. Sustainable fuel cycles are those that improve uranium resource utilization, maximize energy generation, minimize waste generation, improve safety and limit proliferation risk. DOE-NE chartered a Study on the evaluation and screening of nuclear fuel cycle options, to provide information about the potential benefits and challenges of nuclear fuel cycle options and to identify a relatively small number of promising fuel cycle options with the potential for achieving substantial improvements compared to the current nuclear fuel cycle in the United States. The identification of these promising fuel cycles helps in focusing and strengthening the U.S. R&D investment needed to support the set of promising fuel cycle system options and nuclear material management approaches. DOE-NE is developing and evaluating advanced technologies for the immobilization of waste issued from aqueous and electrochemical recycling activities including off-gas treatment and advanced fuel fabrication. The long-term scope of waste form development and performance activities includes not only the development, demonstration, and technical maturation of advanced waste management concepts but also the development and parameterization of defensible models to predict the long-term performance of waste forms in geologic disposal. Along with the finding of the Evaluation and Screening Study will be presented the major research efforts that are underway for the development and demonstration of waste forms and processes including glass ceramic for high-level waste raffinate, alloy waste forms and glass ceramics composites for HLW from the electrochemical processing of fast reactor fuels, and high durability waste forms for radioiodine.

In order to increasing the thermal conductivity of fuel pellet used in nuclear power plants, a UO 2 -BeO composite fuel was developed. The fuel pellets with different beryllia addition were manufactured by sintering and the physical properties were tested and compared with uranium dioxide fuel. The metallograph show that the continuous phase of beryllia was formed. The results show that the thermal conductivity was obviously larger and the elasticity modulus and the coefficient of thermal expansion were a little smaller than uranium dioxide. The thermal conductivity and the elasticity modulus were increased and the coefficient of thermal expansion was decreased by increasing the ratio of beryllia contents.

Nozzles of fuel assembly play an important role in pressure water reactor (PWR) fuel assembly element. For a long time, ordinary processing technologies of nozzles of fuel assembly have the problems of difficult and complicated process, the low availability of material and long the development cycles of manufacturing. However, according to the study these issues can be well settled by using the additive manufacturing technology. This paper studies a nozzle of fuel assembly prepared by this additive manufacturing technology through slow-strain-rate tension (SSRT) test and microstructure observation experiment. The results of SSRT test show that yield strength of the nozzle of fuel assembly is about 401.5MPa, the extensional rigidity is about 673.5MPa and the ductility is about 45.7%. And the SEM fracture results of the SSRT sample indicate that the fracture microstructure contains a large number of dimples, and the way of fracture belongs to plastic. And the metallographic observation consequences manifest that the microstructure of nozzle of fuel assembly prepared by the additive manufacturing technology is composite tissue of both austenite and ferrite, and the grains are settled along the way of laser scanning and there are isometric with some kind of direction. This metallographic microstructure is different from the traditional morphology of the free carbide distributed in the matrix. The dual phase microstructure of austenite and ferrite can improve the mechanical properties of the matrix effectively, and avoid the free carbides which may lead to matrix fragmentation in the tensile deformation process. Moreover, the laser power could affect the microstructure and properties of nozzles of fuel assembly observably, and the high laser power could bring about the ablation of metal. Through the analysis of mechanical properties and microstructure, we have made it possible to make the laser additive manufacturing technology to be used for the fuel assembly nozzle preparation in the nuclear power area. This work not only presents the advantages of the laser additive manufacturing technology in the fuel element processing area of the nuclear power station, but also broadens the application range of the laser additive manufacturing technology. What’s more we provide the new thoughts for the fast and effective preparation of the fuel element especially for the fuel assembly nozzle in the nuclear power station.

In this study, we analyze and compare the temperature profiles, and temperature gradient profiles of UN, U 3 Si 2 and U 3 O 8 aluminum (Al) dispersed nuclear fuels to propose safer nuclear fuels with enhanced thermal conductivity. To calculate the electronic and lattice thermal conductivities, we use EPW, BoltzTrap and ShengBTE codes implemented with Quantum Espresso. Maxwell-Eucken approximation is used to get the effective thermal conductivity of the considered dispersed fuels. The temperature and temperature gradients are calculated by solving the steady state heat conduction equation for a cylindrical fuel rod. Results show that these fuels have reduced the centerline temperature which will prevent fuel melting, as well it will reduce the thermal stress which leads to cracking the pellet.

Silicon carbide (SiC) and SiC matrix composites (SiC f /SiC) are being investigated as potential fuel cladding materials for advanced PWRs in order to improve the safety of nuclear power plants. The conceptual design of multi-layered SiC cladding (consisting of a monolithic SiC layer, SiC f /SiC composite layer and a monolithic SiC coating layer) has been investigated to meet the fuel requirements of both the strength and impermeability. A stress distribution model of the triple-layered SiC is developed on the basis of the theory of thermo-elasticity mechanics, taking radial temperature gradient and swelling effects into account as well. The heat transferring behavior of the cladding is investigated by analyzing the temperature distribution under steady conditions. Finite Element Analysis (FEA) code ANSYS is used to obtain the stress and temperature nephogram of multi-layered SiC fuel cladding under simulated steady conditions. Compared with the results of ANSYS, the stress distribution model and temperature distribution is validated.

SiC fiber composite material as fuel cladding is proposed to increase the power density and maximum allowable fuel burnup in light water reactors. Empirical models about thermal properties of the SiC material are developed as a function of operating temperature and neutron fluence. A fuel rod modeling code frapcon2-SiC based on frapcon2 is compiled to predict the performance of SiC cladding when operating. Comparison of the behavior between the SiC f /SiC cladding and Zr-4 cladding in different thickness reveals that higher temperature will get due to the poor thermal conductivity of the SiC. As far as the Conclusion shows that Poor thermal performance of SiC cladding makes the design margin becomes smaller.

A novel type of flexible composite with tungsten powder filled in polymers (TPFP) homogeneously was developed as an alternative to lead (Pb) based radiation shields used in nuclear industries. TPFP had a density in the range of 4 to 11.3 g/cm 3 , which can be tailor-made according to the applications. In addition to the advantage of lower toxicity over Pb-based shielding, TPFP can be formed into various shapes, such as pipe shields, pipe wraps, safe floor shields, blankets, etc. The mechanical properties and attenuation of γ-ray was investigated for the developed TPFP.

The fire and explosion assessment is necessary for NPP to evaluate the effect caused by malicious large commercial aircraft impact. This report focuses on the fire and explosion effect on shield building structure of nuclear island according to the impact screening. With the investigation and analysis for the fire and explosion of aircraft impact, a simple and effective method was established. With this method the pool fire and fire explosion are analyzed and their effects on shield building structure are assessed based on the temperature and pressure distribution obtained from the FEA model.

Boron Carbide (B 4 C) particle-reinforced aluminum matrix composite is the key material for use as neutron absorber plate in spent fuel storage racks as well as new fuel and in-containment fuel storage racks for GENIII advanced passive nuclear power plants in China. This material has once depended upon importing with high expense and restricted delivery schedule by foreign supplier. Therefore it has meaningful practical significance to realize the localized manufacturing for this material in China. More importantly, since it’s the first time for this material to be used in domestic plant, particular care should be taken to assure the formal supplied neutron absorber material products exhibit high stabilized and reliable service in domestic nuclear engineering. This paper initiates and proposes a principle design framework from technical view in qualification requirements for this neutron absorber material so as to guide the practical engineering application. Aiming at neutron absorber materials supplied under practical manufacturing condition in engineering delivery, the qualification requirements define B 4 C content, matrix chemistry, 10 B isotope, bulk density, 10 B areal density, mechanical property and microstructure as key criteria for material performance. The uniformity assessment as to different locations of this material is also required from at least three lots of material. Only qualified material meeting all of the qualification requirements should proceed to be verified by lifetime testing such as irradiation, corrosion and thermal aging testing. Systematic and comprehensive performance assessments and verification for process stabilization could be achieved through the above qualification. The long-term service for this neutron absorber material in reliable and safe way could be convincingly expected in spent fuel storage application in China.

In order to withstand the malicious impact of large commercial aircraft, the structure of steel concrete is usually used for shield building’s exterior wall, but due to the form of steel concrete structure is special, it often bring difficulties to the computational analysis of finite element in the design and review. The multi-layer shell element is an equivalent method based on the actual physical layering of material, and has the corresponding function in large-scale finite element software (ANSYS, ABAQUS). The method basic idea is to build a shell element based on the mechanics of composite material. According to the requirement, the shell element can be divided into different thicknesses, material properties and appropriate number of integral points. In the finite element calculation, the stress and strain values can be calculated independently for each integration point. This paper aiming the steel concrete structure in the checking calculation of shield building, used the multi-layer shell element method to simulate and analysis. Based on the modal analysis and spectrum analysis calculations of shield building are performed by using the finite element methods, studying the applicable condition of the methods. The analysis showed that the multi-layer shell element method are able to effectively simulate the structure of steel concrete, and can view more detailed analysis of the stress distribution of steel inside the concrete steel and concrete between the layers, in order to better grasp the safety of the shielding plant.

In this paper, we first propose a novel composite nuclear fuel of UO 2 -GaN, which has never been reported before, and then its fully coupled multiphysics fuel performance is investigated using the CAMPUS code developed by ourselves. We propose two different fabrication methods to obtain the UO 2 -GaN fuel, which are Green Granule/Slug Bisque and Spark Plasma Sintering, respectively, resulting in different fuel thermal conductivities. By comparing two kinds of UO 2 -GaN fuel which are fabricated by two methods, we found that fuel fabricated by Green Granule/Slug Bisque possesses high thermal conductivity and performs well during the reactor operation. The gap width, gap conductance, fission gas release, plenum pressure, deviation of oxygen to metal ratio and displacement are all studied in this work. The performance of this novel fuel is also compared with the traditional UO 2 fuel. The UO 2 -GaN enhanced thermal conductivity composite fuel shows the potential of decreasing the fuel temperature, and improving fuel performance and reactor safety. This makes GaN a good candidate to fabricate composite fuel with UO2 from the thermal standpoint. However, this work is to conduct an exploratory approach to the effect for the GaN addition to UO 2 fuel with very limited data. So, further studies are still needed on GaN’s compatibility with UO 2 , neutronic behavior, fission product retention capabilities and irradiation performance, both on experimental measurements and numerical simulations.

Mixed neutron and gamma radiations require different shielding materials as their interaction with materials is different. Composites were developed in order to combine the shielding capabilities of different materials. However, their homogeneity is difficult to be assured which can lead to pinholes where radiation can penetrate. To avoid this problem, several materials arranged in layers can be used to shield against mixed radiations. Since the multilayer shielding can be made from any material in many configurations, the ant colony optimization (ACO) is a promising method because it deals with combinatorial optimization problems. The candidate materials are HDPE, boron, cadmium, gadolinium, tungsten, bismuth, and iron. Preliminary MCNP simulations were done to observe the effect of arrangements, thicknesses, and types of materials on the radiation spectrum. It was found that: (1) the final layer should be made of high density material, (2) an increase beyond certain thicknesses did not result in a significant increase in attenuation, and (3) there should be an optimum combination of material that can effectively shield against both neutrons and gamma rays.

Printed Circuit Heat Exchangers (PCHE) is a new type of compact heat exchangers, it will be widely used for nuclear industry due to its higher heat transfer area density, compact size, and design flexibility. The hydraulic diameter of PCHE tubes ranges from 1mm to 2mm which belongs to mini-channel according to Kandlikar and Grande (2003)’s study. [1] In this paper, two-phase flow boiling heat transfer of water in mini-channel is discussed. The most of previous literatures in this field mainly focused on flow boiling of refrigerants, but the main working fluid in PCHE tubes is water. A composite correlation of flow boiling of water through mini-channel has been developed on basis of a database of water in this paper. Mean absolute error (MAE) method is used to evaluate relative error. Comparing with the experimental data, the MAE of the new correlation is 23.4%.