Robert Sogbadji | University of Ghana (original) (raw)

Papers by Robert Sogbadji

Nuclear Technology and Radiation Protection

In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Gha... more In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Ghana, the Nuclear Power Institute of the Ghana Atomic Energy Commission has on the agenda to advise the government on the nuclear power to include in the country's energy mix. After consideration of several proposed nuclear reactor technologies, the Nuclear Power Institute considered a high pressure reactor or vodo-vodyanoi energetichesky reactor as the nuclear power technologies for Ghana's first nuclear power plant. As part of technology assessments, neutronic safety parameters of both reactors are investigated. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality and neutron behavior at various operating conditions. The high pressure reactor which is still under construction and theoretical safety analysis, showed good inherent safety features which are comparable to the already e...

Environmental Research, Engineering and Management, 2017

The reprocessing of actinides with long half-life has been non-existent except for plutonium (Pu)... more The reprocessing of actinides with long half-life has been non-existent except for plutonium (Pu). This work looks at reducing the actinides inventory nuclear fuel waste meant for permanent disposal. The uranium oxide fuel (UOX) assembly, as in the open cycle system, was designed to reach a burnup of 46GWd/T and 68GWd/T using the MURE code. The MURE code is based on the coupling of a static Monte Carlo code and the calculation of the evolution of the fuel during irradiation and cooling periods. The MURE code has been used to address two different questions concerning the mono-recycling of americium (Am) in present French pressurised water reactors (PWR). These are reduction of americium in the clear fuel cycle and the safe quantity of americium that can be introduced into mixed oxide (MOX) as fuel. The spent UOX was reprocessed to fabricate MOX assemblies, by the extraction of plutonium and addition of depleted uranium to reach burnups of 46GWd/T and 68GWd/T, taking into account various cooling times of the spent UOX assembly in the repository. The effect of cooling time on burnup and radiotoxicity was then ascertained. After 30 years of cooling in the repository, the spent UOX fuel required a higher concentration of Pu to be reprocessed into MOX fuel due to the decay of Pu-241. Americium, with a mean half-life of 432 years, has a high radiotoxicity level, high mid-term residual heat and is a precursor for other long-lived isotopes. An innovative strategy would be to reprocess not only the plutonium from the UOX spent fuel but also the americium isotopes, which presently dominate the radiotoxicity of waste. The mono-recycling of Am is not a definitive solution because the once-through MOX cycle transmutation of Am in a PWR is not enough to destroy all americium. The main objective is to propose a 'waiting strategy' for both Am and Pu in the spent fuel so that they can be made available for further transmutation strategies. The MOX and americium isotopes (MOXAm) fuel was fabricated to see the effect of americium in MOX fuel on the burnup, neutronic behaviour and radiotoxicity. The MOXAm fuel showed relatively good indicators on both burnup and radiotoxicity. A 68GWd/T MOX assembly produced from a reprocessed fuel spent 46GWd/T UOX assembly showed a decrease in radiotoxicity as compared with the open cycle. All fuel types understudied in the PWR cycle showed a good safety inherent feature with the exception of some MOXAm assemblies that have a positive void coefficient in specific configurations, which would not be consistent with safety features.

Nuclear Engineering and Design

journal published by Elsevier. The attached copy is furnished to the author for internal non-comm... more journal published by Elsevier. The attached copy is furnished to the author for internal non-commercial research and education use, including for instruction at the authors institution and sharing with colleagues. Other uses, including reproduction and distribution, or selling or licensing copies, or posting to personal, institutional or third party websites are prohibited. In most cases authors are permitted to post their version of the article (e.g. in Word or Tex form) to their personal website or institutional repository. Authors requiring further information regarding Elsevier's archiving and manuscript policies are encouraged to visit: http://www.elsevier.com/copyright

Journal of Computational Chemistry

Introduction To design a nuclear system properly, it is necessary to predict how the neutrons wil... more Introduction To design a nuclear system properly, it is necessary to predict how the neutrons will be distributed throughout the system. Unfortunately, determining the neutron distribution is a difficult problem in general. The neutrons in a nuclear reactor move in a complicated path as result of repeated nuclear collisions. To a first approximation, however, the overall effect of these collisions is that the neutrons undergo a kind of diffusion in the reactor medium, much like the diffusion of one gas in another. Since neutrons in a nuclear reactor actually have a distribution in energy, this distribution must be accounted for in the diffusion equation. Neutrons are emitted in fission with a continuous energy spectrum, and this distribution broadens as the neutrons are scattered in the medium and diffusion about the system, losing energy in elastic and inelastic collisions. In nuclear engineering, a prompt neutron is a neutron immediately emitted by a nuclear fission event, as oppo...

Nuclear Engineering and Design

The total neutron flux spectrum of the compact core of Ghana's miniature neutron source react... more 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...

Applied Radiation and Isotopes, 2014

ABSTRACT The americium–beryllium neutron irradiation facility at the National Nuclear Research In... more ABSTRACT The americium–beryllium neutron irradiation facility at the National Nuclear Research Institute (NNRI), Ghana, was re-designed with four 20 Ci sources using Monte Carlo N-Particle (MCNP) code to investigate the maximum amount of flux that is produced by the combined sources. The results were compared with a single source Am–Be irradiation facility. The main objective was to enable us to harness the maximum amount of flux for the optimization of neutron activation analysis and to enable smaller sample sized samples to be irradiated. Using MCNP for the design construction and neutronic performance calculation, it was realized that the single-source Am–Be design produced a thermal neutron flux of (1.8±0.0007)×106 n/cm2s and the four-source Am–Be design produced a thermal neutron flux of (5.4±0.0007)×106 n/cm2s which is a factor of 3.5 fold increase compared to the single-source Am–Be design. The criticality effective, keff, of the single-source and the four-source Am–Be designs were found to be 0.00115±0.0008 and 0.00143±0.0008, respectively.

World Journal of Nuclear Science and Technology, 2011

A slightly prompt critical nuclear reactor would increase the neutron flux exponentially at a hig... more 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.

Progress in Nuclear Energy, 2011

ABSTRACT a b s t r a c t The Monte Carlo method was used to determine the neutron fluxes in the i... more ABSTRACT a b s t r a c t The Monte Carlo method was used to determine the neutron fluxes in the irradiation channels of the Ghana Research Reactor-1. The MCNP5 code was used for this purpose to simulate the radial and axial distribution of the neutron fluxes within all the 10 irradiation channels. After the MCNP simulation, it was observed that axially, the fluxes rise to a peak before falling and then finally leveling out. It was also observed that the fluxes were higher in the center of the irradiation channels; the fluxes got higher as it moved toward the center of the core. The multiplication factor (k eff) was observed as 1.000397 AE 0.0007. Radially, the thermal, epithermal and fast neutron flux in the inner irradiation channel range from 1.15 Â 10 12 n/cm 2 .s AE 0.1018 Â 10 11 À 1.19 Â 10 12 n/cm 2 .s AE 0.1172 Â 10 11 , 1.21Â 10 12 n/cm 2 .s AE 0.1014 Â 10 11 À 1.36 Â 10 12 n/cm 2 .s AE 0.1038 Â 10 11 and 2.47 Â 10 11 n/ cm 2 .s AE 0.1120 Â 10 10 À 2.97 Â 10 11 n/cm 2 .s AE 0.1255 Â 10 10 respectively. For the outer channel, the flux range from 7.14 Â 10 11 n/cm 2 .s AE 0.1381 Â 10 10 À 7.38 Â 10 11 n/cm 2 .s AE 0.208 Â 10 10 for thermal, 1.94 Â 10 11 n/cm 2 .s AE 0.1014 Â 10 10 À 2.51 Â 10 11 n/cm 2 .s AE 0.1281 Â 10 10 for epithermal and 3.69 Â 10 10 n/cm 2 .s AE 0.8912 Â 10 8 À 5.14 Â 10 10 n/cm 2 .s AE 0.1009 Â 10 9 for fast. The results have shown that there are flux variations within the irradiation channels both axially and radially.

World Journal of Nuclear Science and Technology, 2011

Accurate neutron flux values in irradiation channels of research reactors are very essential to t... more 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.

This study was conducted to investigate the occurrence and extent of potentially trace and major ... more This study was conducted to investigate the occurrence and extent of potentially trace and major elements in kaolin, widely used in Ghana, using the Instrumental Neutron Activation Analysis (INAA) technique. Soil plays a vital role in human sustenance on earth. Different forms of soil have been used over the years to solve pertinent requirements of man. Kaolin, commonly referred to as "hyire" in Ghana, has been used by women during the delicate periods of their pregnancy and also by lactating mothers. This study has sought to conduct an analysis of the toxic elements that may be ingested by these pregnant women when they eat" hyire" using Instrumental Neutron Activation Analysis (INAA) at the Ghana Research Reactor-1 facility. The study has shown that even though there are useful (major) elements present in the samples used, some toxic elements were also found to be beyond the Recommended Dietary Allowance for those elements. Recommendation has been made to reque...

Nuclear Engineering and Design, 2010

The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate an epicadmiumshield... more The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate an epicadmiumshielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one epicadmium 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 an epicadmium-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 epicadmium-shielded channel was made. The final k eff of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new epicadmium designed model was recorded as 1.00332. Also, a final prompt neutron lifetime of 1.5237 × 10 −4 s was recorded for the new epicadmium designed model while a value of 1.5571 × 10 −7 s was recorded for the original MCNP design of the GHARR-1. The neutron energy causing fission for the original MCNP design of the GHARR-1 was 1.3533 × 10 −2 MeV while that of the new epicadmium designed model was 1.3513 × 10 −2 MeV.

Nuclear Engineering and Design, 2011

The MCNP model for the Ghana Research Reactor-1 (GHARR-1) was redesigned to incorporate cadmiumsh... more The MCNP model for the Ghana Research Reactor-1 (GHARR-1) was redesigned to incorporate cadmiumshielded 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.

Applied Radiation and Isotopes, 2011

The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbideshie... more The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbideshielded 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 Â 10 À 4 s was recorded for the new boron carbide designed model while a value of 1.5571 Â 10 À 7 s was recorded for the original MCNP design of the GHARR-1.

Environmental Research, Engineering and Management, 2011

Biomonitoring of total arsenic and total mercury in the urine of goldmine workers in southwestern... more Biomonitoring of total arsenic and total mercury in the urine of goldmine workers in southwestern Ghana due to occupational exposure was conducted to determine whether occupational exposure substantially contributes to their overall exposure to arsenic and mercury. The urine was collected after 2-day abstinence from sea foods by the workers and from those with no dental amalgam fillings. Total arsenic and total mercury were simultaneously determined by instrumental neutron activation analysis (INAA). After 1-hour irradiation of the urine in Ghana's miniature neutron source reactor (GHARR-1) to induce 76 As and 197 Hg radionuclides through nuclear reactions 75 As(n, γ) 76 As, and 196 Hg(n, γ) 197 Hg, the γ-radiation intensity of the induced 76 As and 197 Hg radionuclides were measured by γ-spectrometry. The validity of the INAA technique for As and Hg determination was checked by analyses of NIST SRM 3103a (As standard solution) and NIST SRM 3133 (Hg standard solution), respectively. The mean mass fractions of arsenic in the urine are 6.76 µg/L ± 1.43, 7.78 µg/L ± 1.33, 8.03 µg/L ± 1.75, 10.44 µg/L ± 1.88, and 14.75 µg/L ± 1.62 for workers in offices 10 km from the mine, 2 km from the mine, 0.5 km from the mine, casual mine workers, and gold ore processing workers, respectively. The levels of arsenic in the urine are all within the 5 to 40 µg As L-1 day-1 normal range for excretion of arsenic. The observed mass fraction of As was higher in high exposure workers. The mean mass fraction of Hg in the urine are 0.36 µg/L ± 0.11, 0.47 µg/L ± 0.12, 0.51 µg/L ± 0.16, 0.57 µg/L ± 0.14, and 0.56 µg/L ± 0.21 for workers in offices 10 km from the mine, 2 km from the mine, 0.5 km from the mine, casual mine workers, and gold ore processing workers, respectively. The high Hg exposed workers engage in small scale gold mining using mercury. The exposure of the different categories of workers to both total arsenic and total mercury are safe.

Annals of Nuclear Energy, 2012

ABSTRACT Evaluation of the flux in various irradiation channels and the absorbed dose rate of rad... more ABSTRACT Evaluation of the flux in various irradiation channels and the absorbed dose rate of radiation by operators 3 cm from the surface of the concrete surrounding the vessel of the 241Am–Be irradiator facility at NNRI, after it had been theoretically modified using MCNP-5 transport code is presented in this work. The 241Am–Be neutron facility at NNRI consist of a single centrally placed neutron source and flanked by two irradiation channels. The theoretical models however had the number of neutron sources increased gradually to five with a fixed number of five irradiation channels of various radii. These were to help assess the viability to increase the irradiator performance and operational capacity. The results indicated an increase in flux in the irradiation channels as well as absorbed dose rates and their dose equivalents as the number of neutron sources increased. The dose equivalent values were all well within the limits set by ICRP and Radiation Protection Board (RPB – Ghana).

Annals of Nuclear Energy, 2011

The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor w... more 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) 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 end of the slowing down region between 8.2491 Â 10 À01 MeV and 8.2680 Â 10 À01 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) Â 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.

Annals of Nuclear Energy, 2011

While there are growing demands for the nuclear data at higher energy regions than keV for up-to-... more While there are growing demands for the nuclear data at higher energy regions than keV for up-to-date scientific and technological development, accurate capture cross sections at thermal energy are still needed. The thermal neutron capture cross sections for the reactions 127 I(n,c) 128 I, 152 Sm(n,c) 153 Sm, 154 Sm(n,c) 155 Sm, and 238 U(n,c) 239 U were determined by the method of foil activation using 55 Mn(n,c) 56 Mn as a reference reaction. The experimental samples with and without a Cd cover were irradiated in an isotropic neutron field of a 20 Ci 241 Am-Be neutron source facility. A high purity Ge detector was used to measure the induced gamma-rays from the samples and the monitor. The thermal neutron capture cross sections of the reactions 127 I(n,c) 128 I, 152 Sm(n,c) 153 Sm, 154 Sm(n,c) 155 Sm, and 238 U(n,c) 239 U were deduced from the analysis of obtained gamma-ray spectra. The thermal neutron capture cross section values for 127 I(n,c) 128 I, 152 Sm(n,c) 153 Sm, 154 Sm(n,c) 155 Sm, and 238 U(n,c) 239 U reactions are (5.93 ± 0.52), (207.3 ± 9.4), (7.7 ± 0.3), and (2.79 ± 0.09) barns respectively. The obtained results have been discussed and compared with the available experimental data and were found to be in agreement with each other.

Annals of Nuclear Energy, 2011

ABSTRACT a b s t r a c t Monte Carlo (MCNP-5) simulations of the neutron fluxes were performed to... more ABSTRACT a b s t r a c t 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 per-centage of the average thermal flux increases as the distance from the source increases, while the per-centages 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.

Nuclear Technology and Radiation Protection

In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Gha... more In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Ghana, the Nuclear Power Institute of the Ghana Atomic Energy Commission has on the agenda to advise the government on the nuclear power to include in the country's energy mix. After consideration of several proposed nuclear reactor technologies, the Nuclear Power Institute considered a high pressure reactor or vodo-vodyanoi energetichesky reactor as the nuclear power technologies for Ghana's first nuclear power plant. As part of technology assessments, neutronic safety parameters of both reactors are investigated. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality and neutron behavior at various operating conditions. The high pressure reactor which is still under construction and theoretical safety analysis, showed good inherent safety features which are comparable to the already e...

Environmental Research, Engineering and Management, 2017

The reprocessing of actinides with long half-life has been non-existent except for plutonium (Pu)... more The reprocessing of actinides with long half-life has been non-existent except for plutonium (Pu). This work looks at reducing the actinides inventory nuclear fuel waste meant for permanent disposal. The uranium oxide fuel (UOX) assembly, as in the open cycle system, was designed to reach a burnup of 46GWd/T and 68GWd/T using the MURE code. The MURE code is based on the coupling of a static Monte Carlo code and the calculation of the evolution of the fuel during irradiation and cooling periods. The MURE code has been used to address two different questions concerning the mono-recycling of americium (Am) in present French pressurised water reactors (PWR). These are reduction of americium in the clear fuel cycle and the safe quantity of americium that can be introduced into mixed oxide (MOX) as fuel. The spent UOX was reprocessed to fabricate MOX assemblies, by the extraction of plutonium and addition of depleted uranium to reach burnups of 46GWd/T and 68GWd/T, taking into account various cooling times of the spent UOX assembly in the repository. The effect of cooling time on burnup and radiotoxicity was then ascertained. After 30 years of cooling in the repository, the spent UOX fuel required a higher concentration of Pu to be reprocessed into MOX fuel due to the decay of Pu-241. Americium, with a mean half-life of 432 years, has a high radiotoxicity level, high mid-term residual heat and is a precursor for other long-lived isotopes. An innovative strategy would be to reprocess not only the plutonium from the UOX spent fuel but also the americium isotopes, which presently dominate the radiotoxicity of waste. The mono-recycling of Am is not a definitive solution because the once-through MOX cycle transmutation of Am in a PWR is not enough to destroy all americium. The main objective is to propose a 'waiting strategy' for both Am and Pu in the spent fuel so that they can be made available for further transmutation strategies. The MOX and americium isotopes (MOXAm) fuel was fabricated to see the effect of americium in MOX fuel on the burnup, neutronic behaviour and radiotoxicity. The MOXAm fuel showed relatively good indicators on both burnup and radiotoxicity. A 68GWd/T MOX assembly produced from a reprocessed fuel spent 46GWd/T UOX assembly showed a decrease in radiotoxicity as compared with the open cycle. All fuel types understudied in the PWR cycle showed a good safety inherent feature with the exception of some MOXAm assemblies that have a positive void coefficient in specific configurations, which would not be consistent with safety features.

Nuclear Engineering and Design

journal published by Elsevier. The attached copy is furnished to the author for internal non-comm... more journal published by Elsevier. The attached copy is furnished to the author for internal non-commercial research and education use, including for instruction at the authors institution and sharing with colleagues. Other uses, including reproduction and distribution, or selling or licensing copies, or posting to personal, institutional or third party websites are prohibited. In most cases authors are permitted to post their version of the article (e.g. in Word or Tex form) to their personal website or institutional repository. Authors requiring further information regarding Elsevier's archiving and manuscript policies are encouraged to visit: http://www.elsevier.com/copyright

Journal of Computational Chemistry

Introduction To design a nuclear system properly, it is necessary to predict how the neutrons wil... more Introduction To design a nuclear system properly, it is necessary to predict how the neutrons will be distributed throughout the system. Unfortunately, determining the neutron distribution is a difficult problem in general. The neutrons in a nuclear reactor move in a complicated path as result of repeated nuclear collisions. To a first approximation, however, the overall effect of these collisions is that the neutrons undergo a kind of diffusion in the reactor medium, much like the diffusion of one gas in another. Since neutrons in a nuclear reactor actually have a distribution in energy, this distribution must be accounted for in the diffusion equation. Neutrons are emitted in fission with a continuous energy spectrum, and this distribution broadens as the neutrons are scattered in the medium and diffusion about the system, losing energy in elastic and inelastic collisions. In nuclear engineering, a prompt neutron is a neutron immediately emitted by a nuclear fission event, as oppo...

Nuclear Engineering and Design

The total neutron flux spectrum of the compact core of Ghana's miniature neutron source react... more 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...

Applied Radiation and Isotopes, 2014

ABSTRACT The americium–beryllium neutron irradiation facility at the National Nuclear Research In... more ABSTRACT The americium–beryllium neutron irradiation facility at the National Nuclear Research Institute (NNRI), Ghana, was re-designed with four 20 Ci sources using Monte Carlo N-Particle (MCNP) code to investigate the maximum amount of flux that is produced by the combined sources. The results were compared with a single source Am–Be irradiation facility. The main objective was to enable us to harness the maximum amount of flux for the optimization of neutron activation analysis and to enable smaller sample sized samples to be irradiated. Using MCNP for the design construction and neutronic performance calculation, it was realized that the single-source Am–Be design produced a thermal neutron flux of (1.8±0.0007)×106 n/cm2s and the four-source Am–Be design produced a thermal neutron flux of (5.4±0.0007)×106 n/cm2s which is a factor of 3.5 fold increase compared to the single-source Am–Be design. The criticality effective, keff, of the single-source and the four-source Am–Be designs were found to be 0.00115±0.0008 and 0.00143±0.0008, respectively.

World Journal of Nuclear Science and Technology, 2011

A slightly prompt critical nuclear reactor would increase the neutron flux exponentially at a hig... more 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.

Progress in Nuclear Energy, 2011

ABSTRACT a b s t r a c t The Monte Carlo method was used to determine the neutron fluxes in the i... more ABSTRACT a b s t r a c t The Monte Carlo method was used to determine the neutron fluxes in the irradiation channels of the Ghana Research Reactor-1. The MCNP5 code was used for this purpose to simulate the radial and axial distribution of the neutron fluxes within all the 10 irradiation channels. After the MCNP simulation, it was observed that axially, the fluxes rise to a peak before falling and then finally leveling out. It was also observed that the fluxes were higher in the center of the irradiation channels; the fluxes got higher as it moved toward the center of the core. The multiplication factor (k eff) was observed as 1.000397 AE 0.0007. Radially, the thermal, epithermal and fast neutron flux in the inner irradiation channel range from 1.15 Â 10 12 n/cm 2 .s AE 0.1018 Â 10 11 À 1.19 Â 10 12 n/cm 2 .s AE 0.1172 Â 10 11 , 1.21Â 10 12 n/cm 2 .s AE 0.1014 Â 10 11 À 1.36 Â 10 12 n/cm 2 .s AE 0.1038 Â 10 11 and 2.47 Â 10 11 n/ cm 2 .s AE 0.1120 Â 10 10 À 2.97 Â 10 11 n/cm 2 .s AE 0.1255 Â 10 10 respectively. For the outer channel, the flux range from 7.14 Â 10 11 n/cm 2 .s AE 0.1381 Â 10 10 À 7.38 Â 10 11 n/cm 2 .s AE 0.208 Â 10 10 for thermal, 1.94 Â 10 11 n/cm 2 .s AE 0.1014 Â 10 10 À 2.51 Â 10 11 n/cm 2 .s AE 0.1281 Â 10 10 for epithermal and 3.69 Â 10 10 n/cm 2 .s AE 0.8912 Â 10 8 À 5.14 Â 10 10 n/cm 2 .s AE 0.1009 Â 10 9 for fast. The results have shown that there are flux variations within the irradiation channels both axially and radially.

World Journal of Nuclear Science and Technology, 2011

Accurate neutron flux values in irradiation channels of research reactors are very essential to t... more 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.

This study was conducted to investigate the occurrence and extent of potentially trace and major ... more This study was conducted to investigate the occurrence and extent of potentially trace and major elements in kaolin, widely used in Ghana, using the Instrumental Neutron Activation Analysis (INAA) technique. Soil plays a vital role in human sustenance on earth. Different forms of soil have been used over the years to solve pertinent requirements of man. Kaolin, commonly referred to as "hyire" in Ghana, has been used by women during the delicate periods of their pregnancy and also by lactating mothers. This study has sought to conduct an analysis of the toxic elements that may be ingested by these pregnant women when they eat" hyire" using Instrumental Neutron Activation Analysis (INAA) at the Ghana Research Reactor-1 facility. The study has shown that even though there are useful (major) elements present in the samples used, some toxic elements were also found to be beyond the Recommended Dietary Allowance for those elements. Recommendation has been made to reque...

Nuclear Engineering and Design, 2010

The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate an epicadmiumshield... more The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate an epicadmiumshielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one epicadmium 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 an epicadmium-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 epicadmium-shielded channel was made. The final k eff of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new epicadmium designed model was recorded as 1.00332. Also, a final prompt neutron lifetime of 1.5237 × 10 −4 s was recorded for the new epicadmium designed model while a value of 1.5571 × 10 −7 s was recorded for the original MCNP design of the GHARR-1. The neutron energy causing fission for the original MCNP design of the GHARR-1 was 1.3533 × 10 −2 MeV while that of the new epicadmium designed model was 1.3513 × 10 −2 MeV.

Nuclear Engineering and Design, 2011

The MCNP model for the Ghana Research Reactor-1 (GHARR-1) was redesigned to incorporate cadmiumsh... more The MCNP model for the Ghana Research Reactor-1 (GHARR-1) was redesigned to incorporate cadmiumshielded 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.

Applied Radiation and Isotopes, 2011

The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbideshie... more The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbideshielded 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 Â 10 À 4 s was recorded for the new boron carbide designed model while a value of 1.5571 Â 10 À 7 s was recorded for the original MCNP design of the GHARR-1.

Environmental Research, Engineering and Management, 2011

Biomonitoring of total arsenic and total mercury in the urine of goldmine workers in southwestern... more Biomonitoring of total arsenic and total mercury in the urine of goldmine workers in southwestern Ghana due to occupational exposure was conducted to determine whether occupational exposure substantially contributes to their overall exposure to arsenic and mercury. The urine was collected after 2-day abstinence from sea foods by the workers and from those with no dental amalgam fillings. Total arsenic and total mercury were simultaneously determined by instrumental neutron activation analysis (INAA). After 1-hour irradiation of the urine in Ghana's miniature neutron source reactor (GHARR-1) to induce 76 As and 197 Hg radionuclides through nuclear reactions 75 As(n, γ) 76 As, and 196 Hg(n, γ) 197 Hg, the γ-radiation intensity of the induced 76 As and 197 Hg radionuclides were measured by γ-spectrometry. The validity of the INAA technique for As and Hg determination was checked by analyses of NIST SRM 3103a (As standard solution) and NIST SRM 3133 (Hg standard solution), respectively. The mean mass fractions of arsenic in the urine are 6.76 µg/L ± 1.43, 7.78 µg/L ± 1.33, 8.03 µg/L ± 1.75, 10.44 µg/L ± 1.88, and 14.75 µg/L ± 1.62 for workers in offices 10 km from the mine, 2 km from the mine, 0.5 km from the mine, casual mine workers, and gold ore processing workers, respectively. The levels of arsenic in the urine are all within the 5 to 40 µg As L-1 day-1 normal range for excretion of arsenic. The observed mass fraction of As was higher in high exposure workers. The mean mass fraction of Hg in the urine are 0.36 µg/L ± 0.11, 0.47 µg/L ± 0.12, 0.51 µg/L ± 0.16, 0.57 µg/L ± 0.14, and 0.56 µg/L ± 0.21 for workers in offices 10 km from the mine, 2 km from the mine, 0.5 km from the mine, casual mine workers, and gold ore processing workers, respectively. The high Hg exposed workers engage in small scale gold mining using mercury. The exposure of the different categories of workers to both total arsenic and total mercury are safe.

Annals of Nuclear Energy, 2012

ABSTRACT Evaluation of the flux in various irradiation channels and the absorbed dose rate of rad... more ABSTRACT Evaluation of the flux in various irradiation channels and the absorbed dose rate of radiation by operators 3 cm from the surface of the concrete surrounding the vessel of the 241Am–Be irradiator facility at NNRI, after it had been theoretically modified using MCNP-5 transport code is presented in this work. The 241Am–Be neutron facility at NNRI consist of a single centrally placed neutron source and flanked by two irradiation channels. The theoretical models however had the number of neutron sources increased gradually to five with a fixed number of five irradiation channels of various radii. These were to help assess the viability to increase the irradiator performance and operational capacity. The results indicated an increase in flux in the irradiation channels as well as absorbed dose rates and their dose equivalents as the number of neutron sources increased. The dose equivalent values were all well within the limits set by ICRP and Radiation Protection Board (RPB – Ghana).

Annals of Nuclear Energy, 2011

The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor w... more 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) 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 end of the slowing down region between 8.2491 Â 10 À01 MeV and 8.2680 Â 10 À01 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) Â 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.

Annals of Nuclear Energy, 2011

While there are growing demands for the nuclear data at higher energy regions than keV for up-to-... more While there are growing demands for the nuclear data at higher energy regions than keV for up-to-date scientific and technological development, accurate capture cross sections at thermal energy are still needed. The thermal neutron capture cross sections for the reactions 127 I(n,c) 128 I, 152 Sm(n,c) 153 Sm, 154 Sm(n,c) 155 Sm, and 238 U(n,c) 239 U were determined by the method of foil activation using 55 Mn(n,c) 56 Mn as a reference reaction. The experimental samples with and without a Cd cover were irradiated in an isotropic neutron field of a 20 Ci 241 Am-Be neutron source facility. A high purity Ge detector was used to measure the induced gamma-rays from the samples and the monitor. The thermal neutron capture cross sections of the reactions 127 I(n,c) 128 I, 152 Sm(n,c) 153 Sm, 154 Sm(n,c) 155 Sm, and 238 U(n,c) 239 U were deduced from the analysis of obtained gamma-ray spectra. The thermal neutron capture cross section values for 127 I(n,c) 128 I, 152 Sm(n,c) 153 Sm, 154 Sm(n,c) 155 Sm, and 238 U(n,c) 239 U reactions are (5.93 ± 0.52), (207.3 ± 9.4), (7.7 ± 0.3), and (2.79 ± 0.09) barns respectively. The obtained results have been discussed and compared with the available experimental data and were found to be in agreement with each other.

Annals of Nuclear Energy, 2011

ABSTRACT a b s t r a c t Monte Carlo (MCNP-5) simulations of the neutron fluxes were performed to... more ABSTRACT a b s t r a c t 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 per-centage of the average thermal flux increases as the distance from the source increases, while the per-centages 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.