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Fig. 4.10. Axial Distribution of the Total Source af Delayed Neutrons in the MSRE with 233U Fuel Loading.
corresponds to the just-cril.ica1, circulating condition.
By comparison, the distribution of the prompt -
neutron source, as well as the delayed-neutron
source in the noncirculating condition, is proportional
to the fission density curve in Fig. 4.6.
The effective (reduced) values of pi, the delayed
fractions for group i, in the critical, circulating
condition are listed in column 4 of Table 4.4. To
obtain these values, the components of the total
delayed-neutron source distribution shown in
Fig. 4.10 must be weighted by the fuel salt volume
fraction and also by the distribution of fast-group
importance shown in Fig. 4.6. These calculations
were made by using an IHM 7090 numerical integration
program, based on the theory given in ref. 6.
The difference between the sums of the pi columns
of Table 4.4 for the stationary and circulating conditions
is equal to the reactivity loss due to circulation.
In normal reactor operation, this difference
will be observed as an additional increment of
regulat.ing rod withdrawal, asymptotically required
to maintain criticality when going from noncirculating
to circulating conditions.
Tuble 4.4. Delayed-Neutron Fractions in the
MSRE with 233U Fuel
1 0.0126 2.28 1.091
2 0.0337 7.8.5 3.548
3 0.139 6.64 4.036
4 0.325 7.36 5.9b2
5 1.130 1.36 1.330
6 2.500 0.85 0.876
Total 26.40 17.14
aG. R. Keepin, Physics of Nuclear Kinetics, p. 90,
Addison-Wesley, Reading, Mass., 1965.
In Fig. 4.11 are curves showing the reactivity
addition requited to produce a given positive stable
period, starting from the critical condition (01 = 0
at zero point of reactivity). Curve A, for t.he non-
circulating condition, is calculated with the
s