- 12 -


angular neutron flux (number of neutrons per unit area per unit time per unit solid angle per unit speed in. the volume
           -4
element dr about r, traveling within the solid angle d in the direction  , and having a speed v in dv at time t)


Zt(r,v)   macroscopic collision cross section for
          neutrons at r with speed v

   X(v)   spectrum of fission neutrons

   v(v) = mean numbers of neutrons resulting from
          a fission caused by a neutron with
          speed v


= macroscopic scattering cross section at r for changing the speed and direction v'1, ' into a speed and direction range dv, dQ at v,


        The individual terms of (2.1) describe the following reactions:

           1      change per unit time in the net number
           v Dt   of neutrons of direction   and speed v
                  in dr


•-  V  =


net loss per unit time from dr of neutrons of direction  in d  by leakage


Et@= number of neutrons of direction , and
      speed v which are removed per unit time
      from dr by absorption and scattering
      collisions


Xfdv'fdQ'V lf =




  fdv'fd-,'Z Sp =


number of neutrons of direction Q2 and speed v gained per unit time in dr from fissions due to neutrons with all speeds v and directions

number of neutrons of direction   and speed v gained per unit time in dr from scattering, collisions which scatter neutrons from all speeds v' and directions


ES (~


(rVA t)