• Energy Research
• 7. Clean energy close
• 15. Life on land close
• Annals of Nuclear Energy close

Abstract In this paper, coupled governing equations were proposed to simulate three-dimensional heat conduction and moisture transport in the nuclear waste repository. Because there will be a large temperature gradient near inner and outer boundaries of the bentonite buffer layer, and the heat driven moisture transport is introduced in the governing equation of moisture transport. By applying the Fourier and Laplace transforms, the governing equations were converted to the Bessel equations, and the Laplace-domain solutions to temperature and moisture distributions were derived through the inverse Fourier transform. The reliability of the proposed solutions was verified through comparative analysis with the line heat source model. These analytical solutions were applied to obtain the evolutions of temperature field and moisture distribution near the waste canister. Finally, a sensitivity study was performed to analyze the effects of relevant parameters on the heat driven moisture transport.

Abstract Molten salt reactors (MSRs) are a class of next-generation nuclear reactors that have received recent industrial and research interest. A generalized species transport solver was implemented in the Virtual Environment for Reactor Applications (VERA) computing suite to extend this tool to analyze liquid-fueled MSRs. This core simulator has been extended to model the transport of fission product gases into a collection of circulating gas bubbles with the purpose of removing the gases. This paper presents the governing species transport equation, along with various nuclear source terms. Development of the source term for phase migration is discussed, along with a simplified interfacial area tracking method. Finally, a case study on a simplified MSR loop is presented in which modeling parameters were varied to assess their impact on gas removal. The steady state results show that parameters such as bubble diameter, gas injection rate and mass transfer coefficient have a low to moderate effect on the fraction of xenon in the core region. Removal efficiency has the greatest effect on the fraction in the core region. After the pump bowl, bubble diameter has a minor effect on the fraction of xenon in the gas void. These results point out that increasing parameters such as mass transfer coefficient, gas injection rate, and removal efficiency drives the xenon into the circulating gas void, while decreasing bubble diameter also drives xenon into the gas void by increasing interfacial area.

Abstract In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. Two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain the fuel integrity. Because of the numerous possible patterns of fuel assemblies in the reactor core, finding the best configuration is so important and challenging. Different techniques for optimization of fuel loading pattern in the reactor core have been introduced by now. In this study, a software is programmed in C# language to find an order of the fuel loading pattern of a VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process launches by considering an initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. The results on a typical VVER-1000 reactor reveal that the method could reach to a pattern with an allowed radial power peaking factor and increases the cycle length 1.1 days, as well.

Abstract In this work, the zeroth order of average current nodal expansion method (ACNEM) is developed for the neutron noise simulation of nuclear reactors core with two dimensional rectangular fuel assemblies. At first, the static calculation is performed for the forward treatment of diffusion equation. Then the forward neutron noise is earned by solving the diffusion equation in the frequency domain using the zeroth order of ACNEM. For the neutron noise calculation in the domain of reactor core, the noise source is considered as an “absorber with variable strength” type i.e. the absorption cross section can be changed in the selected material. In order to evaluate the accuracy of exploited scheme, the neutron noise simulation is performed for two well-known test cases including 2-D LRA BWR and 2-D BIBLIS PWR. For benchmarking purpose, the adjoint noise calculation is done for comparing results with the forward approach using a conventional relation in an elected non-zero frequency. Also the contrast of results is illustrated between the neutron noise in the zero frequency and the corresponding earned static fluxes. Totally, numerical results of problems validate the accuracy of the neutron noise simulation using the proposed method.

Abstract The sensitivity of FCM parameters to the equivalent radius of the RPT model is evaluated by a quantitative method, and six important FCM parameters are identified. Then the relationship between the six parameters and the equivalent radius of the RPT model is analyzed. Finally, a rapid fitting RPT method is proposed. The new method can directly establish the RPT model by a fitting formula without relying on the results of the Monte Carlo high-fidelity model. The numerical results show that the fitting RPT method achieves high precision in most test cases, which demonstrates practicability for the lattice physics calculation of FCM fuel using conventional PWR lattice physics codes.

Abstract In a pressurized water reactor (PWR), the control rod drop analysis is crucial for the reactor safety and the integrity of core structure. Especially under a seismic condition, the horizontal deflections of the reactor system could cause impact between the control rods and the guide thimbles, and then the resulting frictional forces can significantly increase the scram time and intensify mechanical interaction at the absorber rod cladding and spider springs. The CRDAC code is a control rod scram analysis tool developed by China Nuclear Power Technology Research Institute (CNPRI). In a previous work, Liu et al., 2013 , Li et al., 2013 studied the transverse behavior and the fluid-structure interaction of the whole movement system based on this code. Liu calculated the scram time of a 12ft fuel assembly under the Wenchuan earthquake wave. Huang et al. (2014) proposed a flow model to describe the effect of the flow field parameters on the scram time under the operation condition. This paper presents the seismic analysis method based on the CRDAC code which developed by China Nuclear Power Technology Research Institute (CNPRI). Both the scram time and the seismic response have been calculated, such as the maximum pressure in the guide thimble, impact forces to the spider spring, and friction forces to the absorber rod. According to the design requirements of the third-generation nuclear power plants, safety shutdown earthquake (SSE) and operating basis earthquake (OBE) for the horizontal response need to satisfy 0.3 g and 0.15 g, respectively. Therefore, the maximum accelerations of both the input and the initial horizontal deformation of a fuel assembly have been taken into account in this paper. After several sensitive analyses, it is found that the CRDAC code is a powerful tool for the design of control rods and fuel assemblies.

Abstract Nowadays, the coupled codes technique, which consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into system codes, is extensively used for carrying out best estimate (BE) simulation of complex transient in nuclear power plants (NPP). This technique is particularly suitable for transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. Such complex interactions are encountered under normal and abnormal operating conditions of a boiling water reactors (BWR). In such reactors Oscillations may take place owing to the dynamic behavior of the liquid-steam mixture used for removing the thermal power. Therefore, it is necessary to be able to detect in a reliable way these oscillations. The purpose of this work is to characterize one aspect of these unstable behaviors using the coupled codes technique. The evaluation is performed against Peach Bottom-2 low-flow stability tests number 3 using the coupled RELAP5/PARCS code. In this transient dynamically complex neutron kinetics coupling with thermal-hydraulics events take place in response to a core pressure perturbation. The calculated coupled code results are herein assessed and compared against the available experimental data.

Abstract To analyze the behavior of PWRs fuel rods under steady operation conditions, some models such as modified Forsberg-Massih model and TUBRNP model were applied to FROBA-ROD code, which was developed by Xi'an Jiaotong University in 2012. Then more significant fuel rod behaviors could be simulated by FROBA-ROD code, which facilitated the safety analysis of fuel elements especially under accident conditions. To conduct transient thermal calculations and to improve the simulation accuracy, existing thermal models in FROBA were modified with the new fuel rod thermal conductivity model and burnup distribution model. The new models were validated by comparing it with the results of authorized commercial codes FRAPCON-4.0 and ABQUS and by comparing it with IFA-432 Rod 1 experimental benchmark data conducted by Oak Ridge National Laboratory. The relative calculation errors of centerline temperature calculated by modified FROBA-ROD, original FROBA-ROD and FRAPCON were 5.44%, 11.68% and 9.47% separately, which indicated that the modification and update of FROBA-ROD were valid. Modified FROBA-ROD code was applied to perform the behavior analysis of the AP1000 fuel rod. The simulation results indicated that there were enough safety margins for fuel rod behavior of AP1000 at rated operation conditions. The maximum oxidation thickness of cladding appeared at the upper part of the cladding tube in AP1000, which indicated that this area suffered a higher risk of cladding failure. Modified FROBA-ROD code is expected to deal with fuel rod scenarios under accident conditions after some additions of transient models.

Abstract With the rapid rise of installed nuclear power in China, meeting the increasing demands on natural uranium and rationally treating the vast spent fuel are essential issues for the sustainable development of Chinese nuclear power industry. This paper discusses four most potential nuclear fuel cycle modes in China and analyzes the natural uranium requirements under these different fuel cycle modes first based on three development patterns (low-, medium-, and high-speed) of installed nuclear power capacity. Then, an optimization model including natural uranium requirements, spent fuel final disposal amounts and total cost of electricity generation is constructed and optimization problem under two scenarios of reprocessing capacity are solved and results discussed. The annual and cumulative natural uranium requirements under these two scenarios are also calculated. Finally some conclusions are put forward based on the analyses.