The role of neutron importance in bilinear weighted cross sections for burnup evaluation

Author:

Jafarikia S.,Feghhi S.A.H.

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering

Reference35 articles.

1. Bell, G.I., Glasstone, S., 1970. Nuclear Reactor Theory.

2. Briesmeister, J.F., 2000. MCNP–A General Monte Carlo N-Particle Transport Code, Version 4C, Los Alamos National Laboratory Report LA-13709-M.

3. Interface conditions for few-group neutron diffusion equations with flux-adjoint weighted constant;Buslik;Nucl. Sci. Eng.,1968

4. Double adjoint method for determining the contribution of composition to reactivity at different times;Christie;Ann. Nucl. Energy,2013

5. ORIGEN2: a versatile computer code for calculating the nuclide compositions and characteristics of nuclear materials;Croff;Nucl. Technol.,1983

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