·194·2017
5-14Preliminary Conceptual Design of50MW LBE Cooled
Small Modular Fast Reactor
Gao Yucui and Yang Yongwei
reactor pressure vesselSmall and medium-sized modular reactors normally are reactors whose power usually less than300MW,also, the primary circuit loop such as core,heat exchanger,pump are integrated in the pressure vessel,due to its integration and lower power,the small modular reactor could under the standardized assembly line production and then transported to the destination,in addition,because of the lower power and characteristic of LBE,the LBE cooled small fast reactor could realize natural circulation and then have inherent safety.Thus,the small fast reactor could getflexible application such as centralized heating,water desalination and power supply for remote area and marine or space equipment.In view of those,small and medium-sized modular reactors have been a hot area of research these years.
This paper mainly done the preliminary conceptual design of50MW LBE cooled small modular fast reactor use the open source code openmc.
The reactor core is presented in Fig.1,which has72fuel assemblies,36reflector assemblies and42shielding assemblies and with its height of and diameter of for the purpose of powerflatting,fuel assemblies have three types with different enrichment of19.75%,18.06%,17.65%,core lifetime change with the fuel,Figure.2shows that, compare with MOX fuel,the cores use nitride fuel will have harder spectrum and longer lifetime because of the weaker effect of one nitride whose lifetime will be7.9years.
Fig.1(color online)xy section of the core.Fig.2k effvariation with burn up.
Fig.3(color online)Doppler temperature coefficient.Fig.4(color online)Coolant temperature coefficient.
Figures.3and4shows that the core has negative Coolant temperature coefficient and Doppler temperature coefficient
2017·195·In conclusion,we have the small modular fast reactor whose lifetime is7.9years and temperature coefficient is negative,which meet the design criteria,also,other works such as fuel selection to lengthen the lifetime and natural circulation characteristic will be done in further.
5-15Development and Validation of Reactor Depletion
Calculation Code IMPC-Depletion
Zhao Zelong,YangYongwei,Gao Qingyu,Meng Haiyan and Gao Yucui The nuclide depletion calculation is of great significance in the design and research of the reactor core scheme. The change of nuclide content can be described by the Bateman equation[1],and the kernel of reactor burnup calculation is the solution of Bateman equation.The classic calculation program for nuclide depletion has procedures such as ORIGEN.Since the half-life values of nuclide are widely distributed,the norm of fuel consumption coefficient matrix of the nuclides is very large,so it is difficult to numerically solve the equation.
At present,there are two main methods for solving the fuel consumption equation:Translinear Trajectory Analysis(TTA)method[2]and numerical solution method.The TTA method solves the Bateman equation by decomposing the nuclide consumption chains to a set of linear chains and analytical solution of these subchains will give thefinal results.The procedures using TTA method are Serpent,CINDER90and so on.The numerical solution mainly includes the difference method for ordinary differential equations and the matrix method.ALEPH2, MC21and FISPACT program use difference method solving Bateman equation.The matrix exponential method is the main solution method for numerical solution.According to the different ways for solving the burnup matrix, it is divided into Taylor expansion method,Krylov subspace method and Chebyshev rational approximation[3]etc. This method is fast in the calculation speed and has been used in JMCT,RMC,and Serpent code.
In the nuclide depletion calculation of the ADS system,the external coupling of ORIGEN2.1and MCNPX is still used.However,ORIGEN2.1lacks the product data of some important actinicles nuclides and the nuclide database is limited for ADS system.In order to update and replace ORIGEN2.1program and overcome its deficiencies and improve the accuracy of solution,we develop a new depletion calculation code IMPC-Depletion based on TTA and CRAM method.
IMPC-Depletion is developed on the VS2010platform by the C/C++language,the XMLfile format adopt
ed by OpenMC is used as a standard input.The decay and depletion calculation of nuclides can be performed.The code supports thefixedflux and constant power calculation mode.Current version is1.0and the nuclide database of ORIGEN2.1is used for now.In addition,through the Makefile and g++compiler,it can support the use of IMPC-Depletion under Linux system.To verify the correctness of IMPC-Depletion,decay calculation of U233and fixed neutronflux calculation of U235are performed.The results are compared with ORIGEN2.1using the same nuclide database.
The decay calculation results of U233is shown in Table1.The initial content of U233is1g-atom and the decay time is1×106a.Form Table1,we can see that results of IMPC-Depletion show good agreement with ORIGEN2.1. The maximum deviation is about0.06%.So the decay calculation function of IMPC-Depletion is reliable.
Table1Isotopes list for U233decay calculation(unit/g-atom).
Nuclides ORIGEN2.1IMPC-Depletion1.0Deviation/%
U233  6.45774×10−1  6.45967×10−10.0299
Th229  3.13519×10−2  3.13574×10−20.0176
Ac225  1.16961×10−7  1.16981×10−70.0172
Ra225  1.73139×10−7  1.73170×10−70.0179
Fr221  3.89868×10−11  3.89937×10−110.0177
At217  4.37248×10−15  4.37325×10−150.0177
Po213  5.56277×10−19  5.56375×10−190.0177
Bi213  3.70781×10−10  3.70846×10−100.0177
Bi209  3.22878×10−1  3.22675×10−10.0628
Pb209  1.60821×10−9  1.60849×10−90.0175
Tl209  3.85970×10−13  3.86038×10−130.0176
He4  1.96862  1.967410.0615 The constantflux calculation results of U235is shown in Table2.The initial mass of U235is6.0×105,the neutron

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