Facing Challenges for Monte Carlo Analysis of Full PWR Cores : Towards Optimal Detail Level for Coupled Neutronics and Proper Diffusion Data for Nodal Kinetics
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作者:
Nuttin, A.
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Univ Joseph Fourier Grenoble 1, CNRS, IN2P3, LPSC,Inst Polytech Grenoble, Grenoble, FranceUniv Joseph Fourier Grenoble 1, CNRS, IN2P3, LPSC,Inst Polytech Grenoble, Grenoble, France
Nuttin, A.
[1
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Capellan, N.
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Univ Joseph Fourier Grenoble 1, CNRS, IN2P3, LPSC,Inst Polytech Grenoble, Grenoble, FranceUniv Joseph Fourier Grenoble 1, CNRS, IN2P3, LPSC,Inst Polytech Grenoble, Grenoble, France
Capellan, N.
[1
]
David, S.
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机构:
Univ Paris Sud, CNRS, IN2P3, Inst Phys Nucl Orsay, Orsay, FranceUniv Joseph Fourier Grenoble 1, CNRS, IN2P3, LPSC,Inst Polytech Grenoble, Grenoble, France
David, S.
[2
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Doligez, X.
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Univ Paris Sud, CNRS, IN2P3, Inst Phys Nucl Orsay, Orsay, FranceUniv Joseph Fourier Grenoble 1, CNRS, IN2P3, LPSC,Inst Polytech Grenoble, Grenoble, France
Doligez, X.
[2
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El Mhari, C.
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Univ Joseph Fourier Grenoble 1, CNRS, IN2P3, LPSC,Inst Polytech Grenoble, Grenoble, FranceUniv Joseph Fourier Grenoble 1, CNRS, IN2P3, LPSC,Inst Polytech Grenoble, Grenoble, France
El Mhari, C.
[1
]
Meplan, O.
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Univ Joseph Fourier Grenoble 1, CNRS, IN2P3, LPSC,Inst Polytech Grenoble, Grenoble, FranceUniv Joseph Fourier Grenoble 1, CNRS, IN2P3, LPSC,Inst Polytech Grenoble, Grenoble, France
Meplan, O.
[1
]
机构:
[1] Univ Joseph Fourier Grenoble 1, CNRS, IN2P3, LPSC,Inst Polytech Grenoble, Grenoble, France
[2] Univ Paris Sud, CNRS, IN2P3, Inst Phys Nucl Orsay, Orsay, France
PWR;
Monte Carlo;
MCNP;
coupling to thermal-hydraulics;
diffusion data;
basic nodal kinetics;
CANDU REACTOR;
D O I:
10.1051/snamc/201402202
中图分类号:
TP39 [计算机的应用];
学科分类号:
081203 ;
0835 ;
摘要:
Safety analysis of innovative reactor designs requires three dimensional modeling to ensure a sufficiently realistic description, starting from steady state. Actual Monte Carlo (MC) neutron transport codes are suitable candidates to simulate large complex geometries, with eventual innovative fuel. But if local values such as power densities over small regions are needed, reliable results get more difficult to obtain within an acceptable computation time. In this scope, NEA has proposed a performance test of full PWR core calculations based on Monte Carlo neutron transport, which we have used to define an optimal detail level for convergence of steady state coupled neutronics. Coupling between MCNP for neutronics and the subchannel code COBRA for thermal- hydraulics has been performed using the C++ tool MURE, developed for about ten years at LPSC and IPNO. In parallel with this study and within the same MURE framework, a simplified code of nodal kinetics based on two- group and few- point diffusion equations has been developed and validated on a typical CANDU LOCA. Methods for the computation of necessary diffusion data have been defined and applied to NU (Nat. U) and Th fuel CANDU after assembly evolutions by MURE. Simplicity of CANDU LOCA model has made possible a comparison of these two fuel behaviours during such a transient.