atw - International Journal for Nuclear Power | 01.2020
Ever since its first issue in 1956, the atw – International Journal for Nuclear Power has been a publisher of specialist articles, background reports, interviews and news about developments and trends from all important sectors of nuclear energy, nuclear technology and the energy industry. Internationally current and competent, the professional journal atw is a valuable source of information. www.nucmag.com
Ever since its first issue in 1956, the atw – International Journal for Nuclear Power has been a publisher of specialist articles, background reports, interviews and news about developments and trends from all important sectors of nuclear energy, nuclear technology and the energy industry. Internationally current and competent, the professional journal atw is a valuable source of information.
www.nucmag.com
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<strong>atw</strong> Vol. 65 (2020) | Issue 1 ı January<br />
RESEARCH AND INNOVATION 42<br />
So the fast neutron flux and thermal<br />
neutron flux dis tribution on the<br />
coolant change a little.<br />
Thermal conductivity of the MOX (W/m.K)<br />
(16)<br />
Density of the MOX (kg/m 3 )<br />
(17)<br />
Thermal conductivity of the T91 (W/m.K)<br />
(18)<br />
Specific heat capacity of the T91 (J/kg.K)<br />
Thermal conductivity of the Pb (W/m.K)<br />
Density of the Pb (kg/m 3 )<br />
(19)<br />
(20)<br />
(21)<br />
Specific heat capacity of the Pb (J/kg.K)<br />
(22)<br />
Viscosity of the Pb (Pa.s)<br />
(23)<br />
Figure 24, Figure 25, Figure 26 show<br />
the unnormalized fast neutron flux,<br />
unnormalized thermal flux and<br />
temperature distribution on the outer<br />
boundary respectively. Figure 27<br />
shows the fuel pellet centerline<br />
temperature distribution along Z axis<br />
direction and there is the peak<br />
temperature when Z=0.6 m. And the<br />
maximum fuel pellet centerline<br />
temperature deviation is 21 K which<br />
occurs on the outlet. Figure 28 shows<br />
the assembly temperature distribution<br />
along Y axis direction when Z=0.6 m.<br />
Figure 29 shows the coolant temperature<br />
distribution along Z axis<br />
direction. Figure 30 displays the<br />
cladding outer surface temperature<br />
dis tribution along Z axis direction. It<br />
is obvious that the central hole makes<br />
the central fuel pellet temperature<br />
distribution flat, decreases the peak<br />
temperature of fuel pellet and improve<br />
the safety margin. And there is<br />
the maximum fuel temperature in the<br />
central fuel rod of the hot assembly.<br />
The coolant outlet temperature is<br />
755.11 K. The maximum fuel temperature<br />
is 1643.41 K and the maximum<br />
cladding outer surface temperature is<br />
773.83 K calculated by coupling calculation.<br />
And the maximum fuel<br />
temperature is 1647.17 K and the<br />
maximum cladding outer surface<br />
temperature is 777.28 K calculated by<br />
sub-channel code (KMC-SUB) which<br />
are all within the corresponding<br />
thermal- hydraulics design limits.<br />
| Fig. 24.<br />
Fast neutron flux distribution<br />
on the outer boundary.<br />
| Fig. 25.<br />
Thermal neutron flux distribution<br />
on the outer boundary.<br />
| Fig. 27.<br />
The fuel pellet centerline temperature distribution along Z axis direction.<br />
| Fig. 26.<br />
Temperature distribution<br />
on the outer boundary.<br />
6 Conclusion<br />
In this study, based on the UDF and<br />
UDS function of the FLUENT, the<br />
neutron diffusion equation is defined.<br />
There is no requirement to develop<br />
the interface program of the coupling<br />
calculation. Then the assembly is fine<br />
modeled and the solver in the FLUENT<br />
is used to solve the neutron diffusion<br />
equation. The thermal-hydraulics<br />
calculation is carried out at the same<br />
time. There<strong>for</strong>e the coupling calculation<br />
between neutron diffusion<br />
and thermal-hydraulics is achieved on<br />
the same solver of the FLUENT. In<br />
order to achieve the convenient data<br />
transfer, the neutron diffusion and<br />
thermal-hydraulics calculation use<br />
the same meshes.<br />
Through calculating the 2D-TWIGL<br />
benchmark problem by the FLUENT<br />
solver based on the Finite Volume<br />
Method (FVM), and comparing the<br />
effective multiplication factor, neutron<br />
flux and power with reference<br />
values to verify that the solver can<br />
be used to calculate the neutron<br />
diffusion. The errors mainly occur on<br />
Research and Innovation<br />
Research on Neutron Diffusion and Thermal Hydraulics Coupling Calculation based on FLUENT and its Application Analysis on Fast Reactors ı Xuebei Zhang, Chi Wang and Hongli Chen
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