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Neutronic simulation of a European Pressurised Reactor

Neutronic simulation of a European Pressurised Reactor. O.E. Montwedi, V. Naicker School of Mechanical and Nuclear Engineering North-West University. Energy Postgraduate Conference 2013 Cape Town, South Africa. Introduction. Flow chart of the system analysis research. Introduction. EPR

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Neutronic simulation of a European Pressurised Reactor

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  1. Neutronic simulation of a European Pressurised Reactor O.E. Montwedi, V. Naicker School of Mechanical and Nuclear Engineering North-West University Energy Postgraduate Conference 2013 Cape Town, South Africa

  2. Introduction Flow chart of the system analysis research

  3. Introduction • EPR • AREVA NP design GEN III+. • Finland (Olkiluoto ), France (Flamanville), China (2 Units Taishan). • Licencing in USA and UK. • Codes available for neutronic analysis • Diffusion codes >Solve neutron diffusion equation to obtain the neutron flux. >DYN3D neutron kinetics code, NEM (Nodal Expansion Method). • Deterministic codes >Solve the Boltzmann transport equation. > Based mostly on discrete ordinate methods. • Stochastic codes >Uses stochastic methods to simulates particle transport. > Capability to model very complex geometries. > E.g. Monte Carlo N-Particle 5 (MCNP 5).

  4. Introduction Flow chart of the MCNP code for power distribution calculation

  5. Aim • Develop a 3D MCNP EPR Neutronic Model. > Build MCNP input deck. > Establish convergence of the model. • Calculate of the flux and fission powers.

  6. Core model description • Core • 17 X 17 core. • 241 fuel assemblies. • Heavy reflector: stainless steel sheets. • Core barrel :stainless steel. • Moderator : H2O. • RPV : stainless steel. • Assembly • 17x17 Fuel assembly. • 23 Guide tubes. • 1 Instrumentation tube. • 265 Fuel rods.

  7. Results and Discussions • Figures on top 500 n/cycle and 100 000 n/cycle. • Results above shows that the more source neutron per cycles and source point you have the quicker the convergence in both Keff and Source entropy. • If you want reliable results we should discard at least 200 – 300 cycles. B. Brown Forrest, “A review of Monte Carlo criticality calculations–Convergence, Bias, Statistics” Los Alamos National Laboratory (2009).

  8. Results Axial neutron flux distribution (0 0 0) • Flux at the top and bottom of the core is low as expected because of the stainless steel structures present. • Flux at the central region of the core is reduced by incorporation of burnable absorbers(Gd2O3), this increases burn-up.

  9. Results and Discussions Axial power distribution (0 0 0) • As expected the power is distributed similar to the flux distribution. • Power at the central region of the core is reduced by incorporation of burnable absorbers(Gd2O3), this increases burn-up reduce axial leakage.

  10. Conclusion and further work • Model has converged and this allows for stable and reliable results • Multi group flux and fission power is obtained and the distributions follows expected trends. • The results resemble the EPR and PWR design differences and advancements. • Further work • Identification of the hot channel of the core. • Calculation of control rods worth. This work is based upon research supported by the South African Research Chairs Initiative of the Department of Science and Technology and National Research Foundation.

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