JAERI 1287 JNDC Nuclear Data Library of Fission Products Fir
JAERI 1287 JNDC Nuclear Data Library of Fission Products Fir
JAERI 1287 JNDC Nuclear Data Library of Fission Products Fir
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<strong>JAERI</strong><strong>1287</strong> 2. <strong>Nuclear</strong> <strong>Data</strong> <strong>Library</strong> <strong>of</strong> <strong>Fission</strong> <strong>Products</strong> 21<br />
capture reactions <strong>of</strong> fission products. For the other 26 nuclides, fast-neutron 25-group cross<br />
sections were calculated using the data <strong>of</strong> ENDF/B-IV.<br />
For thermal reactor calculations the data in BNL-325 were used for the resonance integral.<br />
Calculated values by <strong>JNDC</strong> were used for the resonance integral when there are no data<br />
in BNL-325. The thermal neutron (2200 m/s) cross sections are evaluated by <strong>JNDC</strong>. For the<br />
thermal reactor systems, the neutron capture rate or the equivalent decay constant X* is<br />
given as the product <strong>of</strong> thermal neutron equivalent cross section and thermal neutron flux by<br />
X*~-oeq-th- (2.6.1)<br />
Thermal neutron equivalent cross section oeq is given by<br />
Oe^A-ozno+r-RI, (2.6.2)<br />
where A is the ratio <strong>of</strong> the average thermal neutron cross section ath for the neutron energies<br />
below 0.5 eV to the cross section a22oo for 2200 m/s neutrons. Therefore, A depends upon<br />
the thermal neutron spectrum or the thermal neutron temperature <strong>of</strong> Tn (K), and is given by<br />
V 4 In<br />
6ff. (2.6.3)<br />
The variable r in Eq. (2.6.2) is the epithermal index or the ratio <strong>of</strong> resonance neutron flux<br />
per unit lethargy to the total thermal neutron flux. RI in Eq. (2.6.2) means the resonance<br />
integral for the neutron capture reaction. For the thermal reactor analysis, the variables r<br />
Table 2.6.2 Structure <strong>of</strong> group constants <strong>of</strong> neutron capture<br />
cross section<br />
Group<br />
1<br />
2<br />
3<br />
4<br />
5<br />
6<br />
7<br />
8<br />
9<br />
10<br />
11<br />
12<br />
13<br />
14<br />
15<br />
16<br />
17<br />
18<br />
19<br />
20<br />
21<br />
22<br />
23<br />
24<br />
25<br />
T<br />
Upper<br />
Energy Range<br />
Lower<br />
10.5 (MeV) 6.5 (MeV)<br />
6.5<br />
4.0<br />
4.0<br />
2.5<br />
2.5<br />
1.4<br />
1.4<br />
0.8<br />
0.8<br />
0.4<br />
0.4<br />
0.2<br />
0.2<br />
0.1<br />
100.0 (KeV) 46.5 (KeV)<br />
46.5<br />
21.5<br />
21.5<br />
10.0<br />
10.0<br />
4.65<br />
4.65<br />
2.15<br />
2.15<br />
1.0<br />
1000.0 (eV) 465.0 (eV)<br />
465.0<br />
215.0<br />
215.0<br />
100.0<br />
100.0<br />
46.5<br />
46.5<br />
21.5<br />
21.5<br />
10.0<br />
10.0<br />
4.65<br />
4.65<br />
2.15<br />
2.15<br />
1.0<br />
1.0<br />
0.465<br />
0.465<br />
0.0252<br />
0.215<br />
Lethargy width<br />
0.480<br />
0.480<br />
0.570<br />
0.570<br />
0.690<br />
0.690<br />
0.690<br />
0.770<br />
0.770<br />
0.770<br />
0.770<br />
0.770<br />
0.770<br />
0.770<br />
0.770<br />
0.770<br />
0.770<br />
0.770<br />
0.770<br />
0.770<br />
0.770<br />
0.770<br />
0.770<br />
0.770<br />
0.770<br />
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