Outer Irradiation Research Articles

Multifunctional Yolk-Shell Mesoporous Silica Obtained via Selectively Etching the Shell: A Therapeutic Nanoplatform for Cancer Therapy.

Herein, we fabricated a new yolk-shell-structured mesoporous silica nanoparticle (YMSN) with multifunctionalities of fluorescence imaging, photothermal therapy (PTT), and drug delivery by using a fluorescein isothiocyanate-doped silica nanoparticle partially covered by patchy gold as the core. Different from the conventional selective etching procedure, the multifunctional silica core is left intact, and the alkali etching mainly occurs in a hexadecyl trimethyl ammonium bromide/silica hybrid layer, which leads to the formation of the void space in YMSNs. In addition, the utilization of patchy gold as the PTT agent can avoid the shield against the outer irradiation on the core. Results show that the as-prepared YMSNs have a good biocompatibility in the concentration of 0-1000 μg·mL-1 and a high doxorubicin-loading capability (8.04 wt %). In vitro and in vivo antitumor experiments reveal that the resulting YMSNs can be utilized for chemo- and photothermic combination therapy as well as optical imaging.

Neutronic assessment of BDBA scenario at the end of Isfahan MNSR core life

Abstract The present study aims to assess the excess induced reactivity in a Miniature Neutron Source Reactor (MNSR) for a Beyond Design Basis Accident (BDBA) scenario. The BDBA scenario as defined in the Safety Analysis Report (SAR) of the reactor involves sticking of the control rod and filling of the inner and outer irradiation sites with water. At the end of the MNSR core life, 10.95 cm of Beryllium is added to the top of the core as a reflector which affects some neutronic parameters such as effective delayed neutrons fraction (β eff ), the reactivity worth of inner and outer irradiation sites that are filled with water and the reactivity worth of the control rod. Given those influences and changes, new neutronic calculations are required to be able to demonstrate the reactor safety. Therefore, a validated MCNPX model is used to calculate all neutronic parameters at the end of the reactor core life. The calculations show that the induced reactivity in the BDBA scenario increases at the end of core life to 7.90 ± 0.01 mk which is significantly higher than the induced reactivity of 6.80 mk given in the SAR of MNSR for the same scenario but at the beginning of the core's life. Also this value is 3.90 mk higher than the maximum allowable operational limit (i.e. 4.00 mk).

A comparative study of the neutron flux spectra in the MNSR irradiation sites for the HEU and LEU cores using the MCNP4C code

A comparative study for fuel conversion from the HEU to LEU in the Miniature Neutron Source Reactor (MNSR) has been performed in this paper using the MCNP4C code. The neutron energy and lethargy flux spectra in the first inner and outer irradiation sites of the MNSR reactor for the existing HEU fuel (UAl4–Al, 90% enriched) and the potential LEU fuels (U3Si2–Al, U3Si–Al, U9Mo–Al, 19.75% enriched and UO2, 12.6% enriched) were investigated using the MCNP4C code. The neutron energy flux spectra for each group was calculated by dividing the neutron flux by the width of each energy group. The neutron flux spectra per unit lethargy was calculated by multiplying the neutron energy flux spectra for each energy group by the average energy of each group. The thermal neutron flux was calculated by summing the neutron fluxes from 0.0 to 0.625eV, the fast neutron flux was calculated by summing the neutron fluxes from 0.5MeV to 10MeV for the existing HEU and potential LEU fuels. Good agreements have been noticed between the flux spectra for the potential LEU fuels and the existing HEU fuels with maximum relative differences less than 10% and 8% in the inner and outer irradiation sites.

Determination of α in the Cadmium Lined Irradiation Channel of the NIRR-1 for Use in <i>k<sub>o</sub></i>-ENAA

The epithermal neutron shape factor (α) was determined in the permanent cadmium lined irradiation channel installed in the large outer irradiation site of the Nigeria Research Reactor-1 (NIRR- 1) by the cadmium covered multimonitor method using the monitors Al-0.1% Au, Zr and Zn thin foils and Mo thin wire and was found to be -0.137 ± 0.018. The high negative value of α indicates a hardened epithermal neutron spectrum in the cadmium lined irradiation channel. The α value obtained was used in the single comparator method of ENAA (ko-ENAA) for the determination of the concentrations of elements in the standard reference material NIST 1515 Apple leaves using Al-0.1% Au thin foil as the single comparator. The concentrations of the elements Sm and Br were determined in the NIST 1515 Apple leaves because of their high Qo values and are in good agreement with the certified values.

An Investigation of the Impact of Installation of Cd-Lined Irradiation Channel in NIRR-1 on Core Physics Data for ENAA

Prior to the installation of the Cd-liner in one of the large outer irradiation channels of NIRR-1, a Monte Carlo simulation was performed using MCNP5 version 1.4 code. This was done to investigate the effect of installation of Cd-liner in either an inner or outer irradiation channel on reactor physics parameters. Data obtained indicate that the core excess reactivity in both inner and outer irradiations channels is reduced by 3.60 ± 0.07 mk and 0.64 ± 0.06 mk, respectively. Considering the fact that NIRR-1 has a cold core excess reactivity of 3.77 mk, results obtained show that installation of the 1 mm thick Cd-sheet in one of the large outer irradiation channels would have no significant impact on the core physics data. After installation of a 1 mm Cd sheath in a large outer irradiation channel, the neutron flux distribution and the stability in the irradiation channels were monitored by foil activation method. Results indicate that the uniformity of neutron flux distribution in the irradiation channel is preserved and the neutron flux data were found to be comparable with the data obtained before the installation.

Assessing the Effect of Fuel Burnup on Control Rod Worth for HEU and LEU Cores of Gharr-1

An important parameter in the design and analysis of a nuclear reactor is the reactivity worth of the control rod which is a measure of the efficiency of the control rod to absorb excess reactivity. During reactor operation, the control rod worth is affected by factors such as the fuel burnup, Xenon concentration, Samarium concentration and the position of the control rod in the core. This study investigates the effect of fuel burnup on the control rod worth by comparing results of a fresh and an irradiated core of Ghana's Miniature Neutron Source Reactor for both HEU and LEU cores. In this study, two codes have been utilized namely BURNPRO for fuel burnup calculation and MCNP5 which uses densities of actinides of the irradiated fuel obtained from BURNPRO. Results showed a decrease of the control rod worth with burnup for the LEU while rod worth increased with burnup for the HEU core. The average thermal flux in both inner and outer irradiation sites also decreased significantly with burnup for both cores.

Design of boron carbide-shielded irradiation site of the outer irradiation site of the Syrian MNSR using MCNP

The MCNP4C model for the Syrian Miniature Neutron Source Reactor (MNSR) was enhanced to incorporate a boron carbide-shielded irradiation site in the first outer irradiation site for a permanent epithermal neutron irradiation site by using boron carbide. The concept of re-designing a new MCNP4C model includes the calculation of the excess core reactivity, control rod worth, shutdown margin and thermal neutron flux in the inner and outer irradiation sites before and after shielding with boron carbide, as well as the calculation of the thermal and epithermal neutron fluxes in the first outer irradiation site. Distribution of the thermal neutron flux in the first outer irradiation site has been calculated using the MCNP4C code and experimentally irradiating five copper wires using the outer irradiation capsule. Good agreement was noticed between the calculated and the measured values. To compensate for the reactivity losses due to the neutrons absorption in the cylindrical boron carbide shell a beryllium shim was added to the top tray of the reactor.

Determination of radial and axial neutron flux distribution in irradiation channel of NIRR-1 using foil activation technique

Recently, one of the outer irradiation channels of the Nigeria Research Reactor-1 (NIRR-1), which is Miniature Neutron Source Reactor (MNSR) was cadmium-lined. In order to ascertain level of flux disturbance on the inner and the remaining outer channels and also to establish flux profiles in sample container as experienced by samples for Neutron Activation Analysis (NAA), a foil activation technique was used to determine the radial and axial neutron flux distribution of the reactor. Measurements were carried out with copper wire flux monitors arranged in radial and axial directions in a typical irradiation container used for routine NAA in NIRR-1. Our results show that for a preset neutron flux of 5.0×1011cm−2s−1 the axial and radial neutron flux of NIRR-1 ranges from 4.47×1011 to 5.16×1011cm−2s−1 and 4.80×1011 to 5.55×1011cm−2s−1 respectively, with mean value of 4.79×1011cm−2s−1 for axial and 4.99×1011cm−2s−1 for radial. The mean value for each position was compared with our previously measured values after commissioning of NIRR-1 (Jonah et al., 2005) and the one done recently at the Ghana MNSR (Abrefah et al., 2010). The values all pointed towards a level of consistency in variation despite the recent installation of cadmium-lined irradiation channel in NIRR-1. This is an indication that the installed cadmium line did not affect NIRR-1 flux stability. However, our individual foils shows slight flux variation from one foil position to another in the same irradiation container as against the claims of MNSR manufacturers and some users for perfect flux stability across the container. There is therefore the need for flux corrections to be made by MNSR users during NAA particularly for samples in the axial position. This can be achieved by sandwiching flux monitors between samples during irradiations.

Determination of k0-factors of short-lived nuclides and application of k0-NAA to selected trace elements

As part of the standardization program of k0-based NAA (k0-NAA) methods at the Dalhousie University SLOWPOKE-2 reactor (DUSR) facility, the k0-factors of 15 analytically important short-lived nuclides (half-life <65s) were determined with respect to gold (197Au). The elemental standards used were prepared mostly from their primary standard solutions. The samples were irradiated in both inner and outer pneumatic sites of the DUSR facility and counted using an HPGe-detector coupled to an ORTEC’s digital gamma-ray spectrometer. The k0-factors determined using both inner and outer irradiation sites were found to be within ±5% with respect to either recommended or literature values in most cases. The Z-score values at 95% confidence level were found to be in the range of ±0.03–1.6. The k0-NAA method was applied to three different NIST standard reference materials (SRMs) and concentrations of six elements, namely Ag, F, Hf, Rb, Sc, and Se were determined using their short-lived nuclides. The concentrations of these elements were also determined by relative NAA method for comparison purposes.

The application of Westcott Formalism k0 NAA method to estimate short and medium lived elements in some Ghanaian herbal medicines complemented by AAS

The epithermal neutron shape factor, α of the inner and outer irradiation sites of the Ghana Research Reactor-1 (GHARR-1) was determined obtaining results of 0.105 for the inner (Channel 1) Irradiation site and 0.020 for the outer (channel 6) irradiation site. The neutron temperatures for the inner and outer irradiation sites were 27°C and 20°C, respectively. The α values used in Westcott Formalism k0 INAA was applied to determine multi elements in 13 Ghanaian herbal medicines used by the Centre for Scientific Research into Plant Medicine (CSRPM) for the management of various diseases complemented by Atomic Absorption Spectrometry. They are namely Mist. Antiaris, Mist. Enterica, Mist. Morazia, Mist. Nibima, Mist. Modium, Mist. Ninger, Mist Sodenia, Mist. Tonica, Chardicca Powder, Fefe Powder, Olax Powder, Sirrapac powder and Lippia Tea. Concentrations of Al, As, Br, K, Cl, Cu, Mg, Mn, Na and V were determined by short and medium irradiations at a thermal neutron flux of 5×1011ncm−2s−1. Fe, Cr, Pb, Co, Ni, Sn, Ca, Ba, Li and Sb were determined using Atomic Absorption Spectrometry (AAS). Ba, Cu, Li and V were present at trace levels whereas Al, Cl, Na, Ca were present at major levels. K, Br, Mg, Mn, Co, Ni, Fe and Sb were also present at minor levels. Arsenic was not detected in all samples. Standard Reference material, IAEA-V-10 Hay Powder was simultaneously analysed with samples. The precision and accuracy of the method using real samples and standard reference materials were evaluated and within ±10% of the reported value. Multivariate analytical techniques, such as cluster analysis (Q-mode and R-mode CA) and principal component analysis (PCA)/factor analysis (FA), have been applied to evaluate the chemical variations in the herbal medicine dataset. All the 13 samples may be grouped into 2 statistically significant clusters (liquid based and powdered herbal medicines), reflecting the different chemical compositions. R-mode CA and PCA suggest common sources for Co, Mg, Fe, Ca, Cr, Ni, Sn, Li and Sb and Na, V, Cl, Mn, Al, Br and K. The PCA/FA identified 3 dominant factors as responsible for the data structure, explaining 84.5% of the total variance in the dataset.

The impact of HEU to LEU conversion of commercial MNSR: Determination of neutron spectrum parameters in irradiation channels of NIRR-1 using MCNP code

Miniature Neutron Source Reactor (MNSR) facilities are equipped with 10 irradiation channels mainly for NAA and limited radioisotope production. A neutronics feasibility study has shown that it would be feasible to convert commercial MNSR from HEU (90.2%) to LEU (∼12.5%) fuel. In this work, the impact of the proposed LEU fuel on neutron spectrum parameters in the irradiation channels has been studied. The MCNP code was used to calculate the neutron flux distribution in 640 neutron energy group structure and the cross section data of monitor reactions also in 640 neutron energy group structure were extracted from ENDF libraries. On the basis of the data, the Cd ratios for Au, Co, Th, and U were computed and used to estimate the thermal-to-epithermal neutron ratio, f, as well as the non-ideality of epithermal neutron spectrum, α. In order to verify the calculations, measured experimental data for the current HEU core are compared with calculations. Results obtained indicate slightly “hardened” neutron spectra distributions in the inner and outer irradiation channels of the proposed LEU core with no significant impact on utilization.

Neutron energy spectrum flux profile of Ghana’s miniature neutron source reactor core

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 irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the region monitored. The thermal neutrons recorded their highest flux in the inner irradiation channel with a peak flux of (1.2068 ± 0.0008) × 10 12 n/cm 2 s, followed by the outer irradiation channel with a peak flux of (7.9166 ± 0.0055) × 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) × 10 11 n/cm 2 s. The peak values of the thermal energy range occurred in the energy range (1.8939–3.7880) × 10 −08 MeV. The inner channel again recorded the highest flux of (1.8745 ± 0.0306) × 10 09 n/cm 2 s at the lower energy end of the slowing down region between 8.2491 × 10 −01 MeV and 8.2680 × 10 −01 MeV, but was over taken by the moderator as the neutron energies increased to 2.0465 MeV. The outer irradiation channel recorded the lowest flux in this region. In the fast region, the core, where the moderator is found, the highest flux was recorded as expected, at a peak flux of (2.9110 ± 0.0198) × 10 08 n/cm 2 s at 6.961 MeV. The inner channel recorded the second highest while the outer channel and annulus beryllium recorded very low flux in this region. The flux values in this region reduce asymptotically to 20 MeV.

Feasibility study of photo-neutron flux in various irradiation channels of Ghana MNSR using a Monte Carlo code

Computer simulation was carried out for photo-neutron source variation in outer irradiation channel, inner irradiation channels and the fission channel of a tank-in-pool reactor, a Miniature Neutron Source Reactor (MNSR) in sub-critical condition. Evaluation of the photo-neutron was done after the reactor has been in sub-critical condition for three month period using Monte Carlo Neutron Particle (MCNP) code. Neutron flux monitoring from the Micro Computer Control Loop System (MCCLS) was also investigated at sub-critical condition. The recorded neutron fluxes from the MCCLS after investigations were used to calculate the power of the reactor at sub-critical state. The computed power at sub-critical state was used to normalize the un-normalized results from the MCNP.

Monte Carlo simulation of core physics parameters of the Syrian MNSR reactor

A 3-D neutronic model for the Syrian Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis using the MCNP-4C code. The continuous energy neutron cross sections were evaluated from the ENDF/B-VI library. This model is used in this paper to calculate the following reactor core physics parameters: the clean cold core excess reactivity, calibration of the control rod and calculation its shut down margin, calibration of the top beryllium shim plate reflector, the axial neutron flux distributions in the inner and outer irradiation positions and calculations of the prompt neutron life time (lp) and the effective delayed neutron fraction (βeff). Good agreements are noticed between the calculated and the measured results. These agreements indicate that the established model is an accurate representation of Syrian MNSR core and will be used for other calculations in the future.

Determination of Neutron Fluxes and Spectrum Shaping Factors in Irradiation Sites of Ghana’S Miniature Neutron Source Reactor (mnsr) by Activation Method After Compensation of Loss of Excess Reactivty

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 0.55eV, Al-0.1% Au wires of negligible thickness were irradiated at 3kW 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.

Delayed Neutrons Energy Spectrum Flux Profile of Nuclear Materials in Ghana’s Miniature Neutron Source Reactor Core

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/cm2?s, followed by the outer irradiation channel with an average flux of (2.4524 0.0049) × 1008 n/cm2?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/cm2?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/cm2?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.

Neutron spectrum parameters in two irradiation channels of the Nigeria research reactor-1 (nirr-1) for use in k0-naa

For the utilization of the NIRR-1 to include the k0-standardization method, the following neutron spectrum parameters in one inner and one outer irradiation channels were determined using the “Cd-ratio for multi-monitor method”: The thermal to epithermal flux ratio, f, and the epithermal flux shape factor, α. These parameters determined in the inner irradiation channel A1,are: α = - 0.046 ± 0.005 and f = 18.4 ± 0.34 and for the outer irradiation channel A2 α = + 0.024 ± 0.002 and f = 49.5 ± 0.96. The results compare well with reported values obtained from other reactor facilities with similar core configuration such as the Slowpoke and Miniature Neutron Source Reactor facilities available in literature.

Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP

The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; 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 boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1.

Comparison of the effects of cadmium-shielded and boron carbide-shielded irradiation channel of the Ghana Research Reactor-1

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 epicadmium 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.

Axial and radial distribution of thermal and epithermal neutron fluxes in irradiation channels of the Ghana Research Reactor-1 using foil activation analysis

The experimental method (foil activation) was used to determine the neutron fluxes in two outer irradiation channels of the Ghana Research Reactor-1. In the experimental procedure, it was observed that the fluxes rise to a peak before falling and then finally leveling out, axially. Axially and radially, it was also observed that the fluxes in the center of the channels were lower than those on the sides. Radially, the fluxes dipped in the center while they increased monotonically towards the sides of the channels. The results have shown that there are flux variations within the irradiation channels at both axial and radial directions.