The Nigeria Research Reactor-1 (NIRR-1) is one of the Commercial Miniature Neutron Source Reactors (MNSRs) sited outside China and scheduled for conversion under the auspices of Reduced Enrichment for Research and Test Reactors (RERTR) program. Since 2006, the reduction in the fuel enrichment of MSNR facilities from greater than 90% HEU cores to less than 20% LEU cores has been embarked upon. Consequently in this work, the physics parameters of three dispersion LEU fuels, which include U 3Si, U 3Si 2, and U 9Mo enriched to 19.75% were determined by the MCNP code to investigate their suitability for the conversion of NIRR-1 to LEU. The following reactor core physics parameters were computed for the LEU fuel options: clean cold core excess reactivity ( ρ ex), control rod (CR) worth, shut down margin (SDM), neutron flux distributions in the irradiation channels and kinetics data ( i.e. effective delayed neutron fraction, β eff and prompt neutron lifetime, l f). Results are compared with experimental and calculated data of the current HEU core and indicate that it would be feasible to use any of the LEU options for the conversion of commercial MNSR in general and NIRR-1 in particular from HEU to LEU.
The Nigeria Research Reactor-1 (NIRR-1) is one of the five commercial MNSR facilities designed by China Institute of Atomic Energy (CIAE) that are sited outside China. First criticality was achieved on 3 February 2004 and has been operated safely [
Three qualified LEU fuels were used to substitute the HEU in the original input deck developed for NIRR-1. An MCNP diagram of the established HEU model of NIRR-1 is depicted in
In the MCNP model, the HEU fueled core of NIRR-1 was created in a three-dimensional, Cartesian coordinate system. The input deck was constructed using detailed engineering drawings of the reactor obtained from the SAR [
Fuel Type/ Enrichment % | Density of Meat/U (g/cc) | Meat diameter (mm) | Clad material/ thickness (mm) | No of fuel pins | |
---|---|---|---|---|---|
1 | HEU-UAl4/90.2% | 3.456/0.92 | 4.30 | Al/0.60 | 347 |
2 | LEU-U3Si/19.75% | 7.394/5.49 | 4.30 | Al/0.60 | 347 |
3 | LEU-U3Si2/19.75% | 6.409/4.42 | 4.74 | Al/0.38 | 347 |
4 | LEU-U9Mo/19.75% | 8.210/5.95 | 4.30 | Al/0.60 | 347 |
the MCNP code version 5 - 1.6.
With the increasing availability of affordable computing devices and taking advantage of parallel features of the MCNP code to speed up the Monte Carlo simulations, a computing cluster made up workstations running Microsoft Windows & Linux (Cent OS) operating systems was set up for this work.
On the windows platform, the default binaries supplied with the code was installed, while for the Linux environment, the binaries were built using the Intel Fortran complier 12.1.5 version. MPICH2 version 1.4.1p1 was chosen as the parallel communication software. MPICH is a high performance portable implementation of the Message Passing Interface standards whose software is distributed under the BSD license and is available on common platforms such as Windows, Linux and Mac OS/X. An IP network is configured for layer 3 communications between all the nodes participating in the calculations and the MCNP jobs are assigned based on each computing node’s capabilities such as the processor speed and number of cores per processor. Only the master node needs to have the complete MCNP code with the data library installed, while the platform specific binaries are installed on the secondary nodes.
Several runs were made with 1/2 a million particles in 400 cycles. Tallies were constructed for the calculation of axial neutron flux distributions in the fuel pins, fission chambers irradiation channels and the slant tube. Similarly, tally cards for power distributions in the different fuel pins and the 10 fuel rings were added. The following neutronics data were calculated and they include clean cold core excess reactivity (ρex), control rod (CR) worth, shut down margin (SDM), neutron flux distributions in the irradiation channels and kinetics data (i.e. effective delayed neutron fraction, βeff and prompt neutron lifetime, lf). Results obtained were benchmarked by calculated data for the HEU and measured data in the final SAR
Results of the core physics neutronics data obtained in this work are compared with calculated and measured data for the current HEU core in
Furthermore, magnitudes of the thermal, epithermal and fast neutron flux in the irradiation channels for the current HEU core and the LEU options are compared in
The calculated thermal neutron flux value of 1.19 × 1012 n/cm2.s for the present HEU core at the inner irradiation site is in close agreement with the nominal value of 1.0 × 1012 n/cm2.s typically displayed on the control console during the steady state operation at full power (i.e. 31 kW). In the case of the outer
HEU-UAl4-347 90.2% Measured | HEU UAl4-347 90.2% Calculated | LEU-U3Si 19.75%-347 Calculated | LEU-U3Si2 19.75%-347 Calculated | LEU-U9Mo 19.75%-347 Calculated | |
---|---|---|---|---|---|
keff rod out | 1.00475 ± 0.00006 | 1.00472 ± 0.00006 | 1.00478 ± 0.00006 | 1.00482 ± 0.00006 | |
keff rod in | 0.99715 ± 0.00006 | 0.99799 ± 0.00006 | 0.99807 ± 0.00006 | 0.99829 ± 0.00006 | |
Clean core excess ρ e x = ( k e f f − 1 ) k e f f reactivity, (mk) | 4.97 | 4.73 | 4.70 | 4.76 | 4.80 |
Control rod worth (mk) ρ w = ( k e f f o u t − k e f f i n ) ( k e f f o u t * k e f f i n ) | 7 | 7.59 | 6.71 | 6.69 | 6.51 |
Shut Down Margin (mk) ρw-ρex | 2.03 | 2.86 | 2.01 | 1.93 | 1.71 |
irradiation channel, the calculated thermal neutron flux is 6.77 × 1011 n/cm2∙s representing approximately 50% of the value obtained in the inner channel; this is in agreement with the manufacturer’s data. As shown in
In order to further assess the suitability of the three fuels for NIRR-1 conversion, kinetics parameters of the fuels used in this work are compared with those of the HEU core and the results are presented in
Results show that the βeff for the three fuels compare well with calculated and measured data for the current HEU core. Similarly, the lf for the three fuels
Fuel type | Thermal (0 - 0.625 eV) n/cm2∙s × 1011 | Epithermal (0.625 eV - 0.825 MeV) n/cm2∙s × 1011 | Fast (0.825 MeV - 20 MeV) n/cm2∙s × 1011 | |||
---|---|---|---|---|---|---|
Inner | Outer | Inner | Outer | Inner | Outer | |
HEU-UAl4 | 11.91 ± 0.01 | 6.77 ± 0.01 | 13.39 ± 0.01 | 1.84 ± 0.01 | 2.77 ± 0.01 | 0.368 ± 0.003 |
LEU-U3Si | 11.06 ± 0.01 | 6.18 ± 0.01 | 13.23 ± 0.01 | 1.80 ± 0.01 | 2.75 ± 0.01 | 0.348 ± 0.003 |
LEU-U3Si2 | 11.05 ± 0.01 | 6.54 ± 0.01 | 13.25 ± 0.01 | 1.89 ± 0.01 | 2.74 ± 0.01 | 0.368 ± 0.003 |
LEU-U9-Mo | 10.46 ± 0.01 | 6.26 ± 0.01 | 12.82 ± 0.01 | 1.83 ± 0.01 | 2.65 ± 0.01 | 0.353 ± 0.003 |
Kinetic Parameters | CIAE (quoted values) | HEU 90% (347) | LEU U3Si 19.75% (347) | LEU U3Si2 19.75% (347) | LEU-U9Mo 19.75% (347) |
---|---|---|---|---|---|
βeff (TOTNU) | 0.0081 | 0.00834 ± 0.00008 | 0.00831 ± 0.00008 | 0.00841 ± 0.00008 | 0.00836 ± 0.00008 |
βeff (KOPTS) | 0.0081 | 0.00849 ± 0.00008 | 0.00840 ± 0.00008 | 0.00843 ± 0.00008 | 0.00844 ± 0.00008 |
lf (KOPTS) (μs) | 81.2 | 56.09 ± 0.09 | 50.33 ± 0.09 | 50.21 ± 0.09 | 48.71 ± 0.09 |
compare well with calculated data for the HEU core but deviate from the manufacturer’s value. In order to ascertain the suitability of the three LEU fuels for NIRR-1 conversion, there is a need to investigate their steady state and transients characteristics in comparison with the current HEU core. Consequently, the reactivity feedback coefficients are desirable for this endeavor and will form the basis of further investigations.
The physics parameters of three dispersion LEU fuels, which include U3Si, U3Si2, and U9Mo enriched to 19.75% have been determined by the MCNP code in order to investigate their suitability for the conversion of NIRR-1 to LEU. Results obtained for the clean cold core excess reactivity (ρex), control rod (CR) worth and the shutdown margin (SDM) of the three LEU fuels compare well with data for the HEU core. Additionally, data for the thermal neutron flux in the irradiation channels of the dispersion fuels indicate a reduction of an average of approximately 10% in the magnitude in comparison with the HEU core. Consequently, for an LEU fueled core, the reactor power level would need to be raised from the current value of 31 kW to 34 kW in order to match the nominal flux level (i.e. of 1 × 1012 n/cm2.s in the inner channel). Furthermore, calculated kinetics data of the three LEU fuels indicate their suitability for NIRR-1 conversion to LEU.
Ibikunle, K., Sadiq, U., Ibrahim, Y.V. and Jonah, S.A. (2018) MCNP Simulation of Physics Parameters of Dispersion Fuels for Conversion of NIRR-1 to LEU. World Journal of Nuclear Science and Technology, 8, 23-29. https://doi.org/10.4236/wjnst.2018.82003