192
C. H. Pyeon
Fig. 7.9 (continued)
(d) Case 4
-60
-40
-20
0
20
40
60
80
100
Value in Eq. (6) [pcm]
JENDL-3.3
ENDF/B-VII.0
JEFF-3.1
Pb-204 Pb-206 Pb-207 Pb-208 Al-27 U-235 U-238 total
Isotope
7.2.3 Bismuth Sample Reactivity Worth
7.2.3.1 Eigenvalue Calculations
Numerical analyses were conducted with the use of the Monte Carlo code MCNP6.1
together with the JENDL-4.0 nuclear data library for transport. For actual experimental analyses, the capability of eigenvalue calculations by MCNP6.1 was useful
in the discussion with the use of JENDL-4.0 for processing important data analyses,
and JENDL-4.0 has already been compared with other nuclear data libraries in a
previous study [4], while demonstrating a reference library.
In the reference core shown in Fig. 7.3, criticality was reached by adjusting the
position of control rod C2 and withdrawing control rods C1 and C3, and safety
rods S4, S5 and S6 from the core; excess reactivity was then deduced from the
combined use of control rod worth of C2 by the rod drop method and its calibration
curve by the positive period method. The measured excess reactivity was attained
within an uncertainty of 3%, and compared with the numerical one as shown in
Table 7.8. Here, effective delayed neutron fraction (β eff ) was attained by MCNP6.1
with JENDL-4.0 in both reference (Al: 798 ± 3 pcm) and test (Bi: 801 ± 3 pcm)
Table 7.8 Comparison
between the results of
measured and calculated
(Eq. (7.2); MCNP6.1) excess
reactivities in reference (Al)
and test (Bi) cores (Ref. [3])
Core
Calculation
[pcm]
Experiment
[pcm]
C/E
Reference (Al)
88 ± 6
87 ± 1
1.01 ± 0.07
Case 1
129 ± 6
143 ± 3
0.90 ± 0.05
Case 2
164 ± 6
165 ± 3
0.99 ± 0.04
Case 3
156 ± 6
163 ± 3
0.96 ± 0.04
Case 4
166 ± 6
171 ± 3
0.97 ± 0.04
(β eff = 798 ± 3 [pcm] and = (3.39 ± 0.01)E-05 [s] by MCNP6.1
with JENDL-4.0; C/E: calculation/experiment)
C. H. Pyeon
Fig. 7.9 (continued)
(d) Case 4
-60
-40
-20
0
20
40
60
80
100
Value in Eq. (6) [pcm]
JENDL-3.3
ENDF/B-VII.0
JEFF-3.1
Pb-204 Pb-206 Pb-207 Pb-208 Al-27 U-235 U-238 total
Isotope
7.2.3 Bismuth Sample Reactivity Worth
7.2.3.1 Eigenvalue Calculations
Numerical analyses were conducted with the use of the Monte Carlo code MCNP6.1
together with the JENDL-4.0 nuclear data library for transport. For actual experimental analyses, the capability of eigenvalue calculations by MCNP6.1 was useful
in the discussion with the use of JENDL-4.0 for processing important data analyses,
and JENDL-4.0 has already been compared with other nuclear data libraries in a
previous study [4], while demonstrating a reference library.
In the reference core shown in Fig. 7.3, criticality was reached by adjusting the
position of control rod C2 and withdrawing control rods C1 and C3, and safety
rods S4, S5 and S6 from the core; excess reactivity was then deduced from the
combined use of control rod worth of C2 by the rod drop method and its calibration
curve by the positive period method. The measured excess reactivity was attained
within an uncertainty of 3%, and compared with the numerical one as shown in
Table 7.8. Here, effective delayed neutron fraction (β eff ) was attained by MCNP6.1
with JENDL-4.0 in both reference (Al: 798 ± 3 pcm) and test (Bi: 801 ± 3 pcm)
Table 7.8 Comparison
between the results of
measured and calculated
(Eq. (7.2); MCNP6.1) excess
reactivities in reference (Al)
and test (Bi) cores (Ref. [3])
Core
Calculation
[pcm]
Experiment
[pcm]
C/E
Reference (Al)
88 ± 6
87 ± 1
1.01 ± 0.07
Case 1
129 ± 6
143 ± 3
0.90 ± 0.05
Case 2
164 ± 6
165 ± 3
0.99 ± 0.04
Case 3
156 ± 6
163 ± 3
0.96 ± 0.04
Case 4
166 ± 6
171 ± 3
0.97 ± 0.04
(β eff = 798 ± 3 [pcm] and = (3.39 ± 0.01)E-05 [s] by MCNP6.1
with JENDL-4.0; C/E: calculation/experiment)