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Core designing of a new type of TVS-2M FAs:neutronics and thermal-hydraulics design basis limits

Frontiers in Energy 2021年第1期15卷 256-278页

作者： Saeed GHAEMI Farshad FAGHIHI Department of Nuclear Engineering School of Mechanical EngineeringShiraz UniversityShiraz 71936-16548Iran Radiation Research Center Shiraz UniversityShiraz 71936Iran Lonizing and Non-Ionizing Radiation Center Shiraz University of Medical ScienceShirazIran

One of the most important aims of this study is to improve the core of the current VVER reactors to achieve more burn-up(or more cycle length)and more intrinsic safety.It is an independent study on the Russian new proposed FAs,called TVS-2M,which would be applied for the future advanced VVERs.Some important aspects of neutronics as well as thermal hydraulics investigations(and analysis)of the new type of Fas are conducted,and results are compared with the standards PWR CDBL.The TVS-2M FA contains gadolinium-oxide which is mixed with UO_(2)(for different Gd densities and U-235 enrichments which are given herein),but the core does not contain BARs.The new type TVS-2M Fas are modeled by the SARCS software package to find the PMAXS format for three states of CZP and HZP as well as HFP,and then the whole core is simulated by the PARCS code to investigate transient conditions.In addition,the WIMS-D5 code is suggested for steady core modeling including TVS-2M FAs and/or TVS FAs.Many neutronics aspects such as the first cycle length(first cycle burn up in terms of MWthd/kgU),the critical concentration of boric acid at the BOC as well as the cycle length,the axial,and radial power peaking factors,differential and integral worthy of the most reactive CPS-CRs,reactivity coefficients of the fuel,moderator,boric acid,and the under-moderation estimation of the core are conducted and benchmarked with the PWR CDBL.Specifically,the burn-up calculations indicate that the 45.6 d increase of the first cycle length(which corresponds to 1.18 MWthd/kgU increase of burnup)is the best improving aim of the new FA type called TVS-2M.Moreover,thermal-hydraulics core design criteria such as MDNBR(based on W3 correlation)and the maximum of fuel and clad temperatures(radially and axially),are investigated,and discussed based on the CDBL.

关键词： TVS-2M FAs core design basis limits VVER-1000 analysis mixture of uranium-gadolinium oxides fuels thermal-hydraulics PARCS WIMS-D5

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Implementation of high-fidelity neutronics and thermal–hydraulic coupling calculations in HNET

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Nuclear Science and Techniques 2022年第11期33卷 120-132页

作者： Yan-Ling Zhu Xing-Wu Chen Chen Hao Yi-Zhen Wang Yun-Lin Xu Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering UniversityHarbin 150001China Institute of Nuclear and New Energy Technology(INET) Tsinghua UniversityBeijing 100084China Purdue University West LafayetteIN 47907USA

To perform nuclear reactor simulations in a more realistic manner,the coupling scheme between neutronics and thermal-hydraulics was implemented in the HNET program for both steady-state and transient conditions.For simplicity,efficiency,and robustness,the matrixfree Newton/Krylov(MFNK)method was applied to the steady-state coupling calculation.In addition,the optimal perturbation size was adopted to further improve the convergence behavior of the MFNK.For the transient coupling simulation,the operator splitting method with a staggered time mesh was utilized to balance the computational cost and accuracy.Finally,VERA Problem 6 with power and boron perturbation and the NEACRP transient benchmark were simulated for analysis.The numerical results show that the MFNK method can outperform Picard iteration in terms of both efficiency and robustness for a wide range of problems.Furthermore,the reasonable agreement between the simulation results and the reference results for the NEACRP transient benchmark verifies the capability of predicting the behavior of the nuclear reactor.

关键词： Coupling calculation High-fidelity neutronics thermal-hydraulics Matrix-free Newton/Krylov method Transient simulation

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A comprehensive CFD combustion model for supercritical CFB boilers

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Particuology 2019年第2期17卷 29-37页

作者： Linjie Xu Leming Cheng JieqiangJi Qinhui Wang Mengxiang Fang Institute for thermal Power Engineering State Key Laboratory of Clean Energy Utilization Zhejiang University Hangzhou 31002China

A combustion model of a large-scale supercritical circulati ng fluidized bed (CFB) boiler was developed for comprehensive computational-fluid-dynamics analysis. The model incorporates gas-solid hydrodynamics, coal combustion, heat transfer on heat exchange surfaces in the furnace, and heat transfer between fumace and working medium in the heat transfer tubes. In simulating the dense and dilute phases in the fumace, the gas-solid hydrodynamics is based on the Euler-Euler model and energy-minimization multiscale drag model. Coal combustion entails evaporation, devolatilization, char combustion, gas homoge neous reaction, and pollutant emission. The coefficient ofheat transfer between gas-solid and the waterwall is estimated using the cluster renewal model, and for radiation, the discrete ordinate model is used. Moreover, thermohydraulic processes in the membrane wall are also in eluded in the heat transfer process. The model was successfully applied in simulations of a 350-MW supercritical CFB boiler. Detailed distributions of solids concentration, oxygen, heat flux, and working medium temperature in the boiler furnace are presented.

关键词： Comprehensive combustion model Computational fluid dynamic thermal-hydraulics Supercritical circulating fluidized bed Large-scale furnace

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Development of a dynamics model for graphite-moderated channel-type molten salt reactor

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Nuclear Science and Techniques 2019年第1期30卷 145-155页

作者： Long He Cheng-Gang Yu Rui-Min Ji Wei Guo Ye Dai Xiang-Zhou Cai Shanghai Institute of Applied Physics Chinese Academy of Sciences CAS Innovative Academy in TMSR Energy System Chinese Academy of Sciences University of Chinese Academy of Sciences

A molten salt reactor(MSR) is one of the six advanced reactor concepts selected by the generation Ⅳ international forum because of its advantages of inherent safety, and the promising capabilities of Th-U breeding and transuranics transmutation. A dynamics model for the channel-type MSR is developed in this work based on a three-dimensional thermal–hydraulic model(3DTH) and a point reactor model. The 3DTH couples a three-dimensional heat conduction model and a one-dimensional single-phase flow model that can accurately consider the heat conduction between different assemblies. The 3DTH is validated by the RELAP5 code in terms of the temperature and mass flow distribution calculation. A point reactor model considering the drift of delayed neutron precursors is adopted in the dynamics model. To verify the dynamics model, three experiments from the molten salt reactor experiment are simulated. The agreement of the experimental data and simulation results was excellent.With the aid of this model, the unprotected step reactivity addition and unprotected loss of flow of the 2 MWt experimental MSR are modeled, and the reactor power and temperature evolution are analyzed.

关键词： Molten salt reactor thermal-hydraulics Point reactor model thermal coupling

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Fusion-Driven Sub-Critical Dual-Cooled Waste Transmutation Blanket:Design and Analysis

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Plasma Science and Technology 2003年第6期5卷 2077-2084页

作者：汪卫华吴宜灿柯严康志诚王红艳黄群英 Institute of Plasma Physics Chinese Academy of SciencesHefei 230031China

The Fusion-Driven Sub-critical System (FDS) is one of the Chinese programs to be further developed for fusion application. Its Dual-cooled Waste Transmutation Blanket (DWTB), as one the most important part of the FDS is cooled by helium and liquid metal, and have the features of safety, tritium self-sustaining, high efficiency and feasibility. Its conceptual design has been finished. This paper is mainly involved with the basic structure design and thermal-hydraulics analysis of DWTB. On the basis of a three-dimensional (3-D) model of radial-toroidal sections of the segment box, thermal temperature gradients and structure analysis made with a comprehensive finite element method (FEM) have been performed with the computer code ANSYS5.7 and computational fluid dynamic finite element codes. The analysis refers to the steady-state operating condition of an outboard blanket segment. Furthermore, the mechanical loads due to coolant pressure in normal operating conditions have been also taken into account. All the above loads have been combined as an input for a FEM stress analysis and the resulting stress distribution has been evaluated. Finally, the structure design and Pb-17Li flow velocity has been optimized according to the calculations and analysis.

关键词： dual-cooled transmutation blanket structure design thermal-hydraulics

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Feasibility of an innovative long-life molten chloride-cooled reactor

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Nuclear Science and Techniques 2020年第4期31卷 1-15页

作者： Ming Lin Mao-Song Cheng Zhi-Min Dai Shanghai Institute of Applied Physics Chinese Academy of SciencesJiading CampusShanghai 201800China University of Chinese Academy of Sciences Beijing 100049China

Molten salt-cooled reactor is one of the six GenIV reactors with promising characteristics including safety,reliability,proliferation resistance,physical protection,economics,and sustainability.In this paper,a small innovative molten chloride-cooled fast reactor(MCCFR)with 30-year core and a target 120-MWt thermal power was presented.For its feasible study,neutronics,thermal-hydraulics,and radiation damage analysis were performed.The key design properties including kinetics parameters,reactivity swing,reactivity feedback coefficients,maximum accumulated displacement per atom(DPA)of reactor pressure vessel(RPV)and fuel cladding,and maximum coolant,cladding,and fuel temperatures were evaluated.The results showed the MCCFR could operate without refueling for 30 years with overall negative reactivity feedback coefficients up the end of its life.During its 30-year life,the excess reactivity was well managed by constantly pulling out the control rods.The maximum accumulated DPA on RPV and fuel cladding were 8.92 dpa and 197.03 dpa,respectively,which are both below the design limits.Similarly,the maximum coolant,cladding and fuel center temperatures were all below the design limits during its entire lifetime.According to these results,the MCCFR core design with long life is feasible.

关键词： Molten salt-cooled reactor Neutronics Radiation damage thermal-hydraulics

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A thermal-Hydraulic Coolant Channel Module (CCM) for Single- and Two-Phase Flow

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Applied Mathematics 2015年第12期6卷 2014-2044页

作者： Alois Hoeld Retired from Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) Munich Germany

A theoretical “drift-flux based thermal-hydraulic mixture-fluid coolant channel model” is presented. It is the basis to a corresponding digital “Coolant Channel Module (CCM)”. This purpose derived “Separate-Region Mixture Fluid Approach” should yield an alternative platform to the currently dominant “Separate-Phase Models” where each phase is treated separately. Contrary to it, a direct procedure could be established with the objective to simulate in an as general as possible way the steady state and transient behaviour of characteristic parameters of single- and/or (now non-separated) two-phase fluids flowing within any type of heated or non-heated coolant channels. Their validity could be confirmed by a wide range of verification and validation runs, showing very satisfactory results. The resulting universally applicable code package CCM should provide a fundamental element for the simulation of thermal-hydraulic situations over a wide range of complex systems (such as different types of heat exchangers and steam generators as being applied in both conventional but also nuclear power stations, 1D and 3D nuclear reactor cores etc). Thereby the derived set of equations for different coolant channels (distinguished by their key numbers) as appearing in these systems can be combined with other ODE-s and non-linear algebraic relations from additional parts of such an overall model. And these can then to be solved by applying an appropriate integration routine. Within the solution procedure, however, mathematical discontinuities can arise. This due to the fact that along such a coolant channel transitions from single- to two-phase flow regimes and vice versa could take place. To circumvent these difficulties it will in the presented approach be proposed that the basic coolant channel (BC) is subdivided into a number of sub-channels (SC-s), each of them being occupied exclusively by only a single or a two-phase flow regime. After an appropriate nodalization of the BC (and thus its SC-s) and after applying a “modified

关键词： Applied Mathematics Non-Linear Partial Differential Equations of First Order thermal-hydraulics of Single- and Two-Phase Flow Separate-Region Mixture-Fluid Model Concept

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thermal-hydraulic Stability Analysis of Nuclear Heating Reactors

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Tsinghua Science and Technology 1996年第1期1卷 23-26页

作者：李金才高祖瑛张作义 Institute of Nuclear Energy Technology. Tsinghua University

The two-phase flow instability that can occur in a natural circulation system is of importance in the design of nuclear heating reactors. The time domain code RETRAN-02 and the frequency domain code NUFREQ were applied to estimate the instability boundary and the effects of such parameters as pressure, inlet resistance and riser height in NHR-5 and an experimental loop. The results of the calculations and the experiments are in good agreement. Nonlinear density wave oscillations were analyzed using the RETRAN-02 code. The theory of nonequilibrium thermodynamics was used to find an explicit criterion to estimate the threshold of the stability. Experimental simulation of the nuclear feedback density wave instability was also carried out in a test loop using. computer controlled electric power.

关键词： nuclear heating reactor (NHR) thermal-hydraulics safety flow instability

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