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9.3 Multi-group diffusion theory  107




                  The factors in the equation are as follows:
                     n(r, E, t)¼neutron density at position, r, energy group, E, and time, t. It is also
                     equal to Φ/v, where v¼average neutron velocity in the group.
                     χ(E, E i )¼fraction of the neutrons that are born in energy group, E, because of
                     fissions in energy group, E i .
                     r¼the position vector.
                     E¼neutron energy group.
                     υ(E i )¼number of neutrons produced per fission caused by neutrons in energy
                     group, E i .
                     β¼total delayed neutron fraction.
                     β j ¼delayed neutron fraction for the j-th delayed neutron group.
                     Ʃ f (r, E i )¼macroscopic fission cross section at position, r, in energy group, E i .
                     Φ(r, E i ,t)¼neutron flux at position, r, for energy group, E i at time, t.
                     S(r, E, t)¼rate of neutrons released from an artificial source in energy group, E,
                     at position, r, at time, t.
                     Ʃ R (r, E i !E)¼macroscopic removal cross section for scattering from energy
                     group, E i , to energy group, E, at position, r.
                     Ʃ a (r, E i )¼macroscopic absorption cross section for energy group, E i at
                     position, r.
                     Ʃ R (r, E)¼macroscopic removal cross section for neutron scattering out of energy
                     group, E, at position, r.
                     L(r, E, t)¼leakage of neutrons from energy group, E, at position, r, at time, t.
                     g j (E)¼fraction of delayed neutrons from precursor group, j, that appear in energy
                     group, E.
                     λ j ¼decay constant for the j-th delayed neutron precursor group.
                     C j (r, t)¼concentration of the j-th delayed neutron precursor group at position, r,
                     and at time, t.
                  Two quantities that have not been encountered in previous discussions need expla-
                  nation. These are the removal cross section, Ʃ R , and the neutron leakage term, L(r,
                  E, t).
                     The removal cross section is the probability that a scattering event in energy
                  group, i, results in delivering the scattered neutron into energy group, j. Reactor
                  physics books [1] provide formulas for the removal cross section. The neutron leak-
                  age term, L, is given by the following:
                                         Lr, E, tÞ ¼  rDr Φ r, E i ,tÞ           (9.3)
                                                        ð
                                          ð
                  D¼the diffusion coefficient¼1/(3Ʃ s ).
                     If the diffusion coefficient is constant, Eq. (9.3) becomes
                                                     2
                                                        ð
                                         Lr, E, tÞ ¼  Dr Φ r, E i ,tÞ            (9.4)
                                          ð
                   2
                  r is the Laplacian operator, shown below for radial and axial dependence.
                                                           2
                                               1 ∂   ∂    ∂
                                            2
                                           r ¼      r   +                        (9.5)
                                               r ∂r  ∂r  ∂z 2
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