JAERI 1287 JNDC Nuclear Data Library of Fission Products Fir

JAERI1287 2. Nuclear Data Library of Fission Products 21
capture reactions of fission products. For the other 26 nuclides, fast-neutron 25-group cross
sections were calculated using the data of ENDF/B-IV.
For thermal reactor calculations the data in BNL-325 were used for the resonance integral.
Calculated values by JNDC were used for the resonance integral when there are no data
in BNL-325. The thermal neutron (2200 m/s) cross sections are evaluated by JNDC. For the
thermal reactor systems, the neutron capture rate or the equivalent decay constant X* is
given as the product of thermal neutron equivalent cross section and thermal neutron flux by
X*~-oeq-th- (2.6.1)
Thermal neutron equivalent cross section oeq is given by
Oe^A-ozno+r-RI, (2.6.2)
where A is the ratio of the average thermal neutron cross section ath for the neutron energies
below 0.5 eV to the cross section a22oo for 2200 m/s neutrons. Therefore, A depends upon
the thermal neutron spectrum or the thermal neutron temperature of Tn (K), and is given by
V 4 In
6ff. (2.6.3)
The variable r in Eq. (2.6.2) is the epithermal index or the ratio of resonance neutron flux
per unit lethargy to the total thermal neutron flux. RI in Eq. (2.6.2) means the resonance
integral for the neutron capture reaction. For the thermal reactor analysis, the variables r
Table 2.6.2 Structure of group constants of neutron capture
cross section
Group
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
25
T
Upper
Energy Range
Lower
10.5 (MeV) 6.5 (MeV)
6.5
4.0
4.0
2.5
2.5
1.4
1.4
0.8
0.8
0.4
0.4
0.2
0.2
0.1
100.0 (KeV) 46.5 (KeV)
46.5
21.5
21.5
10.0
10.0
4.65
4.65
2.15
2.15
1.0
1000.0 (eV) 465.0 (eV)
465.0
215.0
215.0
100.0
100.0
46.5
46.5
21.5
21.5
10.0
10.0
4.65
4.65
2.15
2.15
1.0
1.0
0.465
0.465
0.0252
0.215
Lethargy width
0.480
0.480
0.570
0.570
0.690
0.690
0.690
0.770
0.770
0.770
0.770
0.770
0.770
0.770
0.770
0.770
0.770
0.770
0.770
0.770
0.770
0.770
0.770
0.770
0.770
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