Facing Challenges for Monte Carlo Analysis of Full PWR Cores : Towards Optimal Detail Level for Coupled Neutronics and Proper Diffusion Data for Nodal Kinetics-Reference-Cited by-同舟云学术

Facing Challenges for Monte Carlo Analysis of Full PWR Cores : Towards Optimal Detail Level for Coupled Neutronics and Proper Diffusion Data for Nodal Kinetics

Published:2014 Issue: Volume: Page:
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Container-title:SNA + MC 2013 - Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo
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Author:

Nuttin A.,Capellan N.,David S.,Doligez X.,El Mhari C.,Méplan O.

Abstract

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.

Publisher

EDP Sciences

Link

http://sna-and-mc-2013-proceedings.edpsciences.org/10.1051/snamc/201402202/pdf

Reference22 articles.

1. X-5 Monte Carlo Team, MCNP Version 5 - A General Monte Carlo N-Particle Transport Code, Report LA-UR-03-1987, Los Alamos National Laboratory (2003).

2. Méplan O. et al., MURE, MCNP Utility for Reactor Evolution: couples Monte Carlo transport with fuel burnup and thermal-hydraulics calculations, Data Bank Computer Program Services, OECD Nuclear Energy Agency (2009).

3. Basile D. et al., COBRA-EN, an upgraded version of the COBRA-3C/MIT subchannel code for thermal-hydraulics, Data Bank Computer Program Services, OECD Nuclear Energy Agency (2001).

4. Capellan N. et al., “3D coupling of Monte Carlo neutronics and thermal-hydraulics calculations as a simulation tool for innovative reactors”, Proc. Int. Conf. global 2009.

5. Brown F.B., “Review of Monte Carlo Criticality Calculations Convergence, Bias, Statistics”, Proc. Int. Conf. M&C 2009.