Neutron energy spectrum flux profile of Ghana’s miniature neutron source reactor core (original) (raw)
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Nuclear Engineering and Design
The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor was understudied using the Monte Carlo method. To create small energy groups, 20,484 energy grids were used for the three neutron energy regions: thermal, slowing down and fast. The moderator, the inner irra-diation channels, the annulus beryllium reflector and the outer irradiation channels were the region mon-itored. The thermal neutrons recorded their highest flux in the inner irradiation channel with a peak flux of (1.2068 ± 0.0008) Â 10 12 n/cm 2 s, followed by the outer irradiation channel with a peak flux of (7.9166 ± 0.0055) Â 10 11 n/cm 2 s. The beryllium reflector recorded the lowest flux in the thermal region with a peak flux of (2.3288 ± 0.0004) Â 10 11 n/cm 2 s. The peak values of the thermal energy range occurred in the energy range (1.8939–3.7880) Â 10 À08 MeV. The inner channel again recorded the highest flux of (1.8745 ± 0.0306) Â 10 09 n/cm 2 s at the lower energy en...
World Journal of Nuclear Science and Technology, 2011
A slightly prompt critical nuclear reactor would increase the neutron flux exponentially at a high rate causing the reactor to become uncontrollable, however due to the delayed neutrons, it is possible to leave the reactor in a subcritical state as far as only prompt neutrons are concerned and to also sustain the chain reaction when it is going to die out. The delay neutron flux spectrum of the compact core of the Ghana's miniature neutron source reactor (MNSR) was studied using the Monte Carlo method. 20,484 energy groups combined for all three categories of the energy distribution, thermal, slowing down and fast regions were modeled to create small energy bins. The moderator, the inner irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the regions monitored. The delay thermal neutrons recorded its highest flux in the inner irradiation channel with an average flux of (4.0127 0.0076) × 1008 n/cm 2 ·s, followed by the outer irradiation channel with an average flux of (2.4524 0.0049) × 1008 n/cm 2 ·s. The beryllium reflector recorded the lowest flux in the thermal region. These values of the thermal energy range occurred in the energy range (0 -0.625× 10 -07) MeV. The inner irradiation channel again recorded the highest average flux of (1.2050 ± 0.0501) × 1007 n/cm 2 ·s at the slowing down region in the energy range (0.821 -6.94) MeV. The outer irradiation channel recorded the lowest flux in this region. In the fast energy region, (6.96 -20) MeV, the core, where the moderator is found, the same trend was observed with the inner irradiation channel recording the highest flux at an average flux of (2.0647 ± 0.3260) × 1006 n/cm 2 ·s .The outer irradiation channel recorded the second highest flux while the annulus beryllium reflector recorded very low flux in this region. The final k-effective contribution from only delay neutrons is 0.00834 with the delay neutron fraction being 0.01357 ± 0.00049, hence the Ghana MNSR has good safety inherent feature. Figure.1 MCNP plot of GHARR-1 core configuration showing fuel region (reactor core), inner and outer irradiation channels and annulus beryllium reflector [6].
World Journal of Nuclear Science and Technology, 2011
Accurate neutron flux values in irradiation channels of research reactors are very essential to their usage. The total neutron flux of the Ghana Research Reactor-1 (GHARR-1) was measured after a beryllium reflector was added to its shim to compensate for excess reactivity loss. The thermal, epithermal and fast neutron fluxes were determined by the method of foil activation. The experimental samples with and without a cadmium cover of 1-mm thickness were irradiated in the isotropic neutron field of the irradiation sites of Ghana Research Reactor-1 facility. The induced activities in the sample were measured by gamma ray spectrometry with a high purity germanium detector. The necessary correction for gamma attenuation, thermal neutrons and resonance neutron self-shielding effects were taken into account during the experimental analysis. By defining cadmium cutoff energy of 0.55 eV, Al-0.1% Au wires of negligible thickness were irradiated at 3 kW to determine the neutron fluxes of two irradiation channels, outer channel 7 and inner channel 2 whose Neutron Shaping Factor (α) were found to be (0.037 ± 0.001) and (-0.961 ± 0.034). The neutron flux ratios at the inner irradiation site 2 were found to be, (25.308 ± 3.201) for thermal to epithermal neutrons flux, (0.179 ± 0.021) for epithermal to fast neutrons flux and (4.528 ± 0.524) for thermal to fast neutrons flux, in the outer irradiation site 7, the neutron flux ratios were found to be, (40.865 ± 3.622) for thermal to epithermal neutrons flux, (0.286 ± 0.025) for epithermal to fast neutrons flux and (11.680 ± 1.030) for thermal to fast neutrons flux.
Neutron flux distribution in the irradiation channels of Am-Be neutron source irradiation facility
Annals of Nuclear Energy, 2011
Monte Carlo (MCNP-5) simulations of the neutron fluxes were performed to determine the radial and axial neutron fluxes of the two irradiation sites of the 20 Ci 241 Am-Be neutron irradiation facility at NNRI. The geometry of the 241 Am-Be source as well as the irradiator design, constituted one cylindrical neutron source at the center of a cylindrical barrel with water as moderator. In the far and the near irradiation sites that were 13.1 cm and 4.2 cm, respectively, from the source, the average thermal, epithermal and fast neutron fluxes axially increase exponentially from the bottom and peak at the center of the source 3.0 cm from the bottom of the source and decrease to a very low value at the end of the tube. The percentage of the average thermal flux increases as the distance from the source increases, while the percentages of the epithermal and fast fluxes decrease as the distance from source increases. In the far and near irradiation sites the average radial thermal neutron flux decreases at the rates of 307.02 n cm À2 s À1 and 961.54 n cm À2 s À1 per cm along the diameter, respectively. The average radial, epi-thermal and fast neutron fluxes were fairly uniform along the diameter in the two irradiation sites.
Neutron Flux Variation at the Inner Irradiation Channel of the Nigeria Research Reactor-1 (NIRR-1)
Journal of Technology Innovations in Renewable Energy, 2014
In order to ascertain the level of flux variation in one of the inner irradiation channels of the Nigeria Research Reactor-1 (NIRR-1), the irradiation container used for routine activation analysis was employed with copper wires as flux monitors. Measurements were carried out with these wires arranged in axial direction to determine the thermal neutron flux at selected positions using absolute foil activation method. Our results show that there exists a slight flux variation from one position to another ranging from (4.57 ± 0.24) x 10 11 to (5.20 ± 0.20) x 10 11 cm-2 s-1. Individual foil shows slight flux variation from one position to another in the same irradiation container but they all pointed toward a level of stability in spite of the recent installation of the cadmium lined irradiation channel. The values obtained in this work are in good agreement with the previously measured value of (5.14 ± 0.24) x 10 11 cm-2 s-1 after commissioning of NIRR-1. This shows that the cadmium lined installation does not affect the flux stability. In order to improve the accuracy of neutron activation analysis (NAA) using NIRR-1 facility, there is need for flux corrections to be made by miniature neutron source reactor (MNSR) users during NAA particularly long irradiation, where more than six samples are irradiated simultaneously in the same container.
Serbian Journal of Electrical Engineering, 2004
The procedures for the numerical and experimental determination of the neutron flux in the zones with the strong neutron absorption and leakage are described in this paper. Numerical procedure is based on the application of the SCALE-4.4a code system where the Dancoff factors are determined by the VEGA2DAN code. Two main parts of the experimental methodology are measurement of the activity of irradiated foils and determination of the averaged neutron absorption cross-section in the foils by the SCALE-4.4a calculation procedure. The proposed procedures have been applied for the determination of the neutron flux in the internal neutron converter used with the RB reactor core configuration number 114.
Physics of Particles and Nuclei Letters, 2011
The results of experimental and computer modeling investigations of neutron spectra and fluxes obtained with cold and thermal moderators at the IBR 2 reactor (Joint Institute for Nuclear Research (JINR), Dubna) are presented. These studies are for the YuMO small angle neutron scattering (SANS) spec trometer (IBR 2 beamline 4). The neutron spectra have been measured for two methane cold moderators for the standard configuration of the SANS instrument. The data from both moderators under different condi tions of their operation are compared. The ratio of experimentally determined neutron fluxes of cold and thermal moderators is shown at different wavelengths. Monte Carlo simulations have been carried out to determine the spectra for cold methane and thermal moderators. The results of calculations of the ratio of neutron fluxes of cold and thermal moderators at different wavelengths are demonstrated. In addition, the absorption of neutrons in the air gaps on the way from the moderator to the investigated sample is presented. SANS with the protein apoferritin was done with both cold methane and a thermal moderator and the data were compared. The prospects for the use of a cold moderator for a SANS spectrometer at IBR 2 are dis cussed. The advantages of using the YuMO spectrometer with a thermal moderator with respect to the tested cold moderator are shown.
Annals of Nuclear Energy, 2008
Neutron energy spectrum in Miniature Neutron Source Reactor (MNSR), called Pakistan Research Reactor (PARR-2), is measured employing threshold neutron activation detectors. The calculated neutron spectrum was obtained through modeling the core in detail in three-dimensions employing the transport theory based code WIMS-D/4 and the diffusion theory based code CITATION which was also used as pre-information in the adjustment procedure. A Number of threshold detectors in the form of thin foils are used for spectrum measurements. Gamma activity of irradiated foils was measured with the help of a gamma spectroscopic system consisting of a high efficiency HPGe detector and 8000 channels PC based multi-channel analyzer. STAYNL computer code supplied by International Atomic Energy Agency (IAEA) was used for neutron spectrum adjustment. The group cross-section values and their covariance matrices were derived from the data given in preprocessed cross section libraries in ENDF-6 format of IRDF-90/NMF-G. The comparison between theoretical and experimental work shows good agreement.
Applied Radiation and Isotopes, 2011
Determination of thermal to fast neutron flux ratio ðf fast Þ and fast neutron flux ðf fast Þ is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f fast and subsequently f fast were determined using the absolute method. The f fast ranged from 48 to 155, and the f fast was found in the range 1.03 Â 10 10-4.89 Â 10 10 n cm À 2 s À 1. These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor. & 2011 Elsevier Ltd. All rights reserved. 1. 96 Zr(n, g) 97 Zr/ 97m Nb(743.32 keV) 2. 94 Zr(n, g) 95 Zr(724.20+ 756.73 keV) 3. 197 Au(n, g) 198 Au(411.80 keV) 4. 54 Fe(n, p) 54 Mn(834.83 keV) Contents lists available at ScienceDirect
channel of the Ghana Research Reactor-1
Available online xxx a b s t r a c t The MCNP model for the Ghana Research Reactor-1 (GHARR-1) was redesigned to incorporate cadmium-shielded irradiation channel as well as boron carbide-shielded channel in one of the outer irradiation channels. Further investigations were made after initial work in the cadmium-shielded channel to consider the boron carbide-shielded channel and both results were compared to determine the best material for the shielded channel. Before arriving at the final design of only one shielded outer irradiation channel extensive investigations were made into several other possible designs; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model which has a shielded channel is to equip GHARR-1 with the means of performing efficient epithermal neutron activation analysis. The use of epithermal neutron activation analysis can be very useful in many experiments and projects (e.g. it can be used to determine uranium and thorium in sediment samples). After the simulation, a comparison of the results from the boron carbide-shielded channel model for the GHARR-1 and the epicadmium-shielded channel was made. The inner irradiation channels of the two designs recorded peak values of approximately 1.18 × 10 12 ± 0.0036 n/cm 2 s, 1.32 × 10 12 ± 0.0036 n/cm 2 s and 2.71 × 10 11 ± 0.0071 n/cm 2 s for the thermal , epithermal and fast neutron flux, respectively. Likewise the outer irradiation channels of the two designs recorded peak values of approximately 7.36 × 10 11 ± 0.0042 n/cm 2 s, 2.53 × 10 11 ± 0.0074 n/cm 2 s and 4.73 × 10 10 ± 0.0162 n/cm 2 s for the thermal, epithermal and fast neutron flux, respectively. The epi-cadmium design recorded a peak thermal flux of 7.08 × 10 11 ± 0.0033 n/cm 2 s and an epithermal flux of 2.09 × 10 11 ± 0.006 n/cm 2 s in the irradiation channel where the shield was installed. Also, the boron carbide design recorded no peak thermal flux but an epithermal flux of 1.18 × 10 11 ± 0.0079 n/cm 2 s in the irradiation channel where the shield was installed. The final multiplication factor (k eff) of the boron carbide-shielded channel model for the GHARR-1 was recorded as 1.00282 ± 0.0007 while that of the epicadmium designed model was recorded as 1.00332 ± 0.0007. Also, a final prompt neutron lifetime of 1.5237 × 10 −4 ± 0.0008 s was recorded for the cadmium designed model while a value of 1.5245 × 10 −4 ± 0.0008 s was recorded for the boron carbide-shielded design of the GHARR-1.