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Gao M.,Harbin Engineering University | Cao X.,Harbin Engineering University | Zhu C.,State Nuclear Power Software Development Center | Wang J.,Harbin Engineering University
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2017

DRAGON code can conduct 3D method of characteristics (MOC) heterogeneous core calculations by introducing a lot of approximations and multiple acceleration methods, but the computing memory and time are still unaffordable for present engineering application. The transport calculation code DRAGON_HEU which is developed based on the 2D MOC calculation module of DRAGON code employs a coupled planar MOC solution and axial diffusion solution scheme, 2D full core heterogeneous pin MOC calculation and 1D homogeneous axial pin diffusion calculation are coupled through radial and axial leakage under the 3D coarse mesh finite difference (CMFD) global framework, in such way the whole core Pin-by-Pin calculation is achieved. DRAGON_HEU code is testified with the C5G7 3D extension benchmark. Numerical results demonstrate that DRAGON_HEU code can achieve good accuracy while saving a large amount of time. © 2017, Editorial Board of Journal of Nuclear Power Engineering. All right reserved.


Gao M.-M.,Harbin Engineering University | Cao X.-R.,Harbin Engineering University | Zhu C.-L.,State Nuclear Power Software Development Center | Wang J.-M.,Harbin Engineering University
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | Year: 2017

Several acceleration methods for 2D method of characteristics (MOC) are encoded in DRAGON code. However, the coarse mesh finite difference (CMFD) acceleration method with high speedup is not applied. In order to accelerate the 2D MOC calculation in DRAGON with higher speedup, the CMFD acceleration calculation module was developed and its stabilization was also studied. The C5G7-2D benchmark was used to testify the CMFD acceleration module. Numerical results demonstrate that the CMFD acceleration calculation module is more efficient than the other acceleration calculation modules encoded in the original DRAGON code, and is more stable than the OpenMOC code. © 2017, Editorial Board of Atomic Energy Science and Technology. All right reserved.


Li Z.,Shanghai JiaoTong University | Li J.,State Nuclear Power Software Development Center | Lin M.,Shanghai JiaoTong University | Yang Y.,Shanghai JiaoTong University
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2014

An ex-vessel steam explosion is a fuel coolant interaction process which may occur when the reactor vessel fails and the molten core pours into the water in the reactor cavity during a severe reactor accident. A strong enough steam explosion in a nuclear power plant could endanger the containment integrity and lead to a direct release of radioactive material to the environment. In this article, a nuclear island geometrical model of AP1000 nuclear power plant was established and different scenarios of ex-vessel steam explosions in AP1000 NPP were simulated by MC3D code. Since the initial parameters with large quantity of uncertainties under accident condition may have important effects on the steam explosion, some initial parameters study were performed by varying the location of the melt release(75,°45°,30°,0°), the cavity water subcooling, the triggering time for explosion calculations, the melt temperature and the break size. Results indicate that the higher the melt temperature, the longer the triggered time and the lower the coolant temperature would lead to the more severe steam explosion. Besides, when the angle of break reaches 45 degree and the diameter of the break is 0.5m, the steam explosion causes the largest damage. Copyright © 2014 by ASME.


Chou Q.,State Nuclear Power Software Development Center | Liang H.,Tsinghua University | Bai F.,Tsinghua University
Match | Year: 2015

The permanental polynomial of IPR fullerene C70 (D5h) is computed with quadruple precision arithmetic based on sparse graph on PC in acceptable time. The computing adopts 128 bits to store one oat and works well for C70, while the largest fullerene computed before is C60, which can be easily obtained now. Some properties of the coefficients and zeroes of the permanental polynomials of IPR fullerenes C60 and C70 are also investigated. Computational results show the quadruple precision method can handle permanental polynomial of C70 and even larger fullerenes, which are of interest in applications. © 2015 MATCH Commun. Math. Comput. Chem.


Chen M.-L.,China Nuclear Power Technology Research Institute | Lin J.-M.,China Nuclear Power Technology Research Institute | Bai W.,State Nuclear Power Software Development Center
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | Year: 2016

A three-dimensional computational fluid dynamic code CYCAS was developed for hydrogen safety analysis. The code solved three-dimensional, compressible and unsteady Navier-Stokes equations by the numerical method of implicit continuous Eulerian-arbitrary Lagrangian Eulerian. Multi-component mass conservation equations were solved to describe the diffusion and mixing phenomenon. A homogeneous equilibrium model was applied to model the phase change of water vapor in the fluid flow. For the phase change near the wall, the Chilton-Colburn empirical analogy was used. Algebraic and k-ε turbulence models were employed to model the turbulent flow. The turbulent jet experiment HYJET and the international standard problem ISP23 were simulated and analyzed for CYCAS validation. The computational results agree well with the experimental data. © 2016, Editorial Board of Atomic Energy Science and Technology. All right reserved.


Wang L.,Chongqing University | Zeng Z.,Chongqing University | Zhang L.,Chongqing University | Xie H.,Chongqing University | And 2 more authors.
Applied Thermal Engineering | Year: 2016

In a three-distribution-function framework, a lattice Boltzmann model was proposed for thermal flows through porous media under local thermal non-equilibrium conditions, in which both viscous dissipation and compression work were considered. In this model, a density distribution function was used to simulate the flow field and two total energy distribution functions were employed to simulate temperature fields for both fluid and solid. Discrete equilibrium density and total energy distribution functions were obtained from Hermite expansions of the corresponding moment equations. Through Chapman-Enskog procedure, macroscopic governing equations could be recovered expediently from the present thermal lattice Boltzmann model. The proposed model was validated by numerical simulations of thermal Poiseuille flow in a planar porous channel and natural convection in a square porous cavity. © 2016 Elsevier Ltd


Zhang Y.,North China Electrical Power University | Lu D.,North China Electrical Power University | Wu G.,North China Electrical Power University | Du Z.,State Nuclear Power Software Development Center
International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015 | Year: 2015

The heat transfer capacity of Passive Residual Heat Removal Heat Exchanger (PRHR HX) and thermal stratification in the In-containment Refueling Water Storage Tank (IRWST) are of great importance for the efficient and safe removal of the residual heat in the API 000 reactor. The C-type heating rods bundle is being used in the PRHR HX, yet the heat transfer characteristics of this special shape heat exchanger are not completely explicit. In the present work, an overall scaled IRWST (irregular shaped tank, approximate 4m >< 1.5m x 2.5m) and PRHR HX models were built to simulate the thermal-hydraulic process in the residual heat removal accident, which was the first overall scaled separate effect IRWST&PRHR HX experiment compared to the previous work. More than 150 T-type thermocouples were utilized to measure the temperature, and the Particle Image Velocimetry (PIV) were utilized for the measurement of the flow velocity. Based on the experimental data, the transient heat transfer characteristics of C-shaped heating rod bundle were analyzed. Combination factors including the flow resistance, buoyancy-induced flow velocity, and turbulent mixing effects imposed important impacts on the heat transfer capability of the PRHR HX model. The Nu variations trend indicated that the experimental data were in satisfactory agreement with the empirical correlations in the single convection stage, and the heat transfer capability in the vertical section was better.


Zhang Y.,North China Electrical Power University | Lu D.,North China Electrical Power University | Du Z.,State Nuclear Power Software Development Center | Fu X.,State Nuclear Power Software Development Center | Wu G.,North China Electrical Power University
Annals of Nuclear Energy | Year: 2015

The heat transfer effect of Passive Residual Heat Removal Heat Exchanger (PRHR HX) and buoyancy-induced flow in the In-containment Refueling Water Storage Tank (IRWST) are of great importance for the efficient and safe removal of the residual heat in the AP1000 reactor. Although some numerical studies have been conducted, only the standard k-ε model has been applied. Experimental validation of the simulation results was also not sufficient because of the lack of appropriate experimental data. In the present work, the applicability of different Reynolds Average Navier-Stokes (RANS) turbulence models and Large Eddy Simulation (LES) were examined, utilizing the commercial CFD software CFX 14.5. Further, two types of grids were built for the high/low-Reynolds turbulence models, and the y+ values as well as grids sensitivity were carefully analyzed. Meanwhile, overall scaled IRWST and PRHR HX models were built to simulate the thermal-hydraulic process in the residual heat removal accident, which was a new overall scaled separate effect IRWST&PRHR HX experiment. More than 150 thermocouples were utilized to measure the temperature in the key regions, and Particle Image Velocimetry (PIV) was utilized for the measurement of the flow velocity. Based on the validation of turbulence models in simulating the overall variations of temperature and velocity field in the IRWST model, the transient heat transfer capacity of PRHR HX was then analyzed. The results indicated that the low-Reynolds Shear Stress Transport (SST) model with multi-sublayer grid was appropriate for the simulation of buoyancy-induced flow. Nusselt numbers obtained from numerical simulations, experimental data, and empirical correlations were further compared to analyze the heat transfer mechanism. Combination factors including the special C-shape, flow resistance, and turbulent mixing effects imposed important influences on the heat transfer effects of the PRHR HX model. It was confirmed by numerical results, experimental data, as well as empirical correlations that the heat transfer capability of the vertical section was better than the horizontal section. © 2015 Elsevier Ltd. All rights reserved.


Wang S.,State Nuclear Power Software Development Center | Yu H.,State Nuclear Power Software Development Center | Chen Y.-X.,State Nuclear Power Software Development Center | Liu Z.-Q.,State Nuclear Power Software Development Center
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | Year: 2013

The point kinetics calculation module in COSINE software package was introduced, and the realization of the engineering design of point kinetics codes was described. On selection of point kinetics method, the explicit method with stiffness confinement method by the results of various typical transient problems was compared, and the described methods in this paper can fully meet the requirements of reliably design. On design of reactivity model, the idea of comprehensively coverage of every reactivity disturbance introduction forms and every reactivity feedback forms were introduced. In addition, the built-in thermal model was also introduced. The design of kinetic module in this paper has good reliability and flexibility.


Liu M.,State Nuclear Power Software Development Center
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2013

In this paper, the influence of significant nuclear events on public attitude, and the reaction, i.e., public attitude on the development of nuclear industry have been studied. Though it has been widely known that nuclear is a type of clean, green and effective source of energy, due to the characteristics of nuclear, for most common people, they seldom pay attention to the nuclear power plants (NPPs) or the entire nuclear industry except NPPs are built near the residence of their own, or when some catastrophic failures occur. This fact leads to that when the public's attention is attracted to nuclear, the effects on their attitude to nuclear are often negative. Even if there is positive news about nuclear, in most cases, the public will still be worried and prudent. That is one of the reasons why though the nuclear power related techniques has been developing rapidly, in some countries the usage of nuclear power is still quite limited. In order to carry the development of nuclear power forward, to improve the public acceptance is as important as to improve nuclear related science and technology. This paper focuses on how the significant events related to nuclear influence the public acceptance, which will have direct or indirect effects on the development and/or policy of nuclear industry in a country, even the whole world. Additionally, this paper discusses possible and proper solutions to improve the public acceptance to NPPs and nuclear related techniques. Copyright © 2013 by ASME.

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