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.     

Neutron flux determination in irradiation sites of an Am–Be neutron source at NNRI

Neutron flux measurements and flux distribution parameters for two irradiation sites of an Am–Be neutron source irradiator were measured by using gold (Au), zirconium (Zr) and aluminum (Al) foils. thermal neutron flux Φ_th = 1.46 × 10⁴ n cm⁻² s⁻¹ ± 0.01 × 10², epithermal neutron flux Φ_epi = 7.23 × 10² n cm⁻² s⁻¹ ± 0.001, fast neutron flux Φ_f = 1.26 × 10² n cm⁻² s⁻¹ ± 0.020, thermal-to-epithermal flux ratio f = 20.5 ± 0.36 and epithermal neutron shaping factor α = −0.239 ± 0.003 were found for irradiation Site-1; while the thermal neutron flux Φ_th = 4.45 × 10³ n cm⁻² s⁻¹ ± 0.06, the epithermal neutron Φ_epi = 1.50 × 10² n cm⁻² s¹ ± 0.003, the fast neutron flux Φ_f = 1.17 × 10 n cm⁻² s⁻¹ ± 0.011, thermal-to-epithermal flux ratio f = 29.6 ± 0.94, and epithermal neutron shaping factor α = 0.134 ± 0.001 were found for irradiation Site-2. It was concluded that the Am–Be neutron source can be used for neutron activation analysis (NAA). The Am–Be source can be used for neutron activation analysis thereby reducing the burden on GHARR-1 and increasing the research output of the nation.     

Measurement of neutron capture cross-section of the Ga(n, γ)Ga reaction at 0.0536 eV energy

The neutron capture cross-section for the ⁷¹Ga(n,  γ)⁷²Ga reaction at 0.0536 eV energy was measured using activation technique based on TRIGA Mark-II research reactor. The ¹⁹⁷Au(n, γ)¹⁹⁸Au monitor reaction was used to determine the effective neutron flux. Neutron absorption and γ-ray attenuation in gallium oxide pellet were corrected in determination of cross-section. The cross-section for the above reaction at 0.0536 eV amounts to 2.75 ± 0.14 b. As far as we know there are no experimental data available at our investigated energy. So far we are the first, who carried out experiment with 0.0536 eV neutrons for cross-section measurement. The present result is larger than that of JENDL-3.3, but consistent within the uncertainty range. The value of ENDF/B-VII is higher than this work. The result of this work will be useful to observe energy dependence of neutron capture cross-sections.     

Measurements of the Y(n,γ)Y cross-section in the neutron energy range of 13.5-14.6 MeV

The ⁸⁹Y(n,γ)^90mY cross-section has been measured at three neutron energy points between 13.5 and 14.6 MeV using the activation technique and a coaxial HPGe γ-ray detector. The data for the ⁸⁹Y(n,γ)^90mY cross-sections are reported to be 0.39 ± 0.02, 0.43 ± 0.02, and 0.38 ± 0.02 mb at 13.5 ± 0.2, 14.1 ± 0.1, and 14.6 ± 0.2 MeV incident neutron energies, respectively. The first data for the ⁸⁹Y(n,γ)^90mY reaction at neutron energy points of 13.5 and 14.1 MeV are presented. The natural high-purity Y₂O₃ powder was used as target material. The fast neutrons were produced by the T(d,n)⁴He reaction. Neutron energies were determined by the method of making cross-section ratios of ⁹⁰Zr(n,2n)^89m+gZr and ⁹³Nb(n,2n)^92mNb reactions, and the neutron fluencies were determined using the monitor reaction ⁹³Nb(n,2n)^92mNb. The results obtained are compared with existing data.     

Phantom studies of Cd, Hg and Cl by prompt gamma neutron activation analysis using a Pu-Be neutron source

Prompt gamma neutron activation analysis is a means of non-invasive monitoring for occupational exposure to toxic heavy metals such as Cd and Hg. Preliminary kidney detection limits from previous phantom studies at McMaster were 13.6 ± 0.2 ppm for Cd (125 mL phantom) and 315 ± 24 ppm for Hg (125 mL phantom) using the ²³⁸Pu-Be neutron source and 0.88 ± 0.01 ppm for Cd (125 mL phantom) and 16.91 ± 0.05 ppm for Hg (30 mL phantom) using the thermal neutron beam port at the McMaster Nuclear Reactor. The detection limits vary greatly between the two methods due to differences in experimental set-up, neutron energy spectra and a difference in dose by more than a factor of 100. The Hg detection limit from preliminary data is much higher than expected for both neutron source types. In order to explain the apparent detection limit discrepancy, measurements of Hg and Cd phantoms were performed using the ²³⁸Pu-Be neutron source. The results were compared to phantom measurements of Cl, a well-known neutron activation element.     

Suppression of carbon buildup and lifetime improvement by heating carbon stripper foils

We have successfully developed a new method to reduce the amount of carbon buildup on thin cluster (less than 3.5 μg/cm²) carbon stripper foils by heating them with infrared radiation during beam bombardment. We studied the carbon buildup and the foil temperature on foil lifetime using a 2.0 ± 0.5 μA beam of 3.2-MeV Ne⁺ ions. It was found that the carbon buildup begins to rapidly suppress at 460 °C; further, at a foil temperature higher than approximately 820 °C, the initial foil thickness did not change until the foil ruptured. We also found that the carbon buildup shortens the lifetime of stripper foils.The foils treated by the newly developed present method could withstand the maximum and average total beam charges of 530 mC/cm² and 340 mC/cm², respectively, which are approximately 18 and 11 times larger than the values for the best commercially available foils and approximately 3 and 2 times greater than the values for the cluster foils that are not treated by this method.     

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) × 10¹² n/cm² s, followed by the outer irradiation channel with a peak flux of (7.9166 ± 0.0055) × 10¹¹ n/cm² s. The beryllium reflector recorded the lowest flux in the thermal region with a peak flux of (2.3288 ± 0.0004) × 10¹¹ n/cm² 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) × 10⁰⁹ n/cm² 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) × 10⁰⁸ n/cm² 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.     

Determination of thermal neutron cross-section and resonance integral for the As (n, γ)As reaction by activation method using Mn (n, γ)Mn as a monitor

Nuclear constants for use in reactor activation analysis especially (n, γ) cross-sections and absolute gamma intensities, are known to show a rather large scatter in literature. Thermal and resonance cross-sections for the ⁷⁵As (n, γ)⁷⁶As reaction is determined by the method of foil activation using ⁵⁵Mn (n, γ)⁵⁶Mn as a reference reaction. The experimental sample with and without a cadmium cover of 1-mm wall thickness was irradiated in the isotropic neutron field of the outer irradiation sites 7 of Ghana Research Reactor-1 facility which is a miniature neutron source reactor designed by the Chinese. The irradiation channel used has a neutron spectral parameter (α) found to be (0.037 ± 0.001). The induced activity in the sample was measured by gamma ray spectrometry with a high purity germanium detector. A standard solution of Arsenic was used for the analysis. The necessary correction for gamma attenuation, thermal neutrons and resonance neutron self-shielding effects were not taken into account during the experimental analysis because they were negligible. By defining cadmium cut-off energy of 0.55 eV, the result for ⁷⁵As (n, γ)⁷⁶As reaction was found to be: thermal neutron cross-section σ₀ = (4.28 ± 0.19) b and resonance integral I₀ = (61.88 ± 1.07) b.     

Monte Carlo simulation of the thermal column and beam tube of the TRIGA Mark II research reactor

The Monet Carlo simulation of the TRIGA Mark II research reactor core has been performed employing the radiation transport computer code MCNP5. The model has been confirmed experimentally in the PhD research work at the Atominstitute (ATI) of the Vienna University of Technology. The MCNP model has been extended to complete biological shielding of the reactor including the thermal column, radiographic collimator and four beam tubes. This paper presents the MCNP simulated results in the thermal column and one of the beam tubes (beam tube A) of the reactor. To validate these theoretical results, thermal neutron flux density measurements using the gold foil activation method have been performed in the thermal column and beam tube A (BT-A). In the thermal column, the theoretical and experimental results are in fairly good agreement i.e. maximum thermal flux density in the centre decreases in radial direction. Further, it is also agreed that thermal flux densities in the lower part is greater than the upper part of the thermal column. In the BT-A experiment, the thermal flux density distribution is measured using gold foil. The experimental and theoretical diffusion lengths have been determined as 10.77 cm and 9.36 cm respectively with only 13% difference, reflecting good agreement between the experimental and simulated results. To save the computational cost and to incorporate the accurate and complete information of each individual Monte Carlo MC particle tracks, the surface source writing capability of MCNP has been utilized to the TRIGA shielding model. The variance reduction techniques have been applied to improve the statistics of the problem and to save computational efforts.     

Design of the thermal neutron beam for neutron radiography at the Syrian MNSR

Neutron beam design was studied at the Syrian reactor (MNSR, 30 kW) with a view to generating thermal neutron beam in the vertical irradiation sites for neutron radiography. The design of the neutron collimator was performed using MCNP4C and the ENDF/B-V cross-section library. Thermal, epithermal and fast neutron energy ranges were selected as <0.4 eV, 0.4 eV–10 keV, >10 keV, respectively. To produce a good neutron beam quality, bismuth was used as photon filter. In this design, the L/D ratio of this facility had the value of 125. The thermal neutron flux at the beam exit was about 2.548 × 10⁵ n/cm² s. If such neutron beam were built into the Syrian MNSR many scientific applications would be available using the neutron radiography.     

Characterization of the neutron flux distribution at the Athens Tandem Accelerator NCSR “Demokritos”

Quasi-monoenergetic neutron beams, in the energy range 7-11.5 MeV, produced via the ²H(d,n) reaction, have been used at the 5.5 MV tandem T11/25 Accelerator Laboratory of NCSR “Demokritos”. The flux variation of the neutron beam is monitored with a BF₃ detector, while the absolute flux is obtained with respect to reference reactions. An investigation of the energy dependence of the neutron fluence has been carried out using two independent techniques: by a liquid scintillator BC501A detector and deconvolution of its recoil energy spectra performed by means of the DIFBAS code, as well as via the multiple foil activation technique in combination with the SULSA unfolding code. The neutron facility has also been characterized by means of Monte Carlo simulations with MCNP5.     

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 × 10¹² ± 0.0036 n/cm² s, 1.32 × 10¹² ± 0.0036 n/cm² s and 2.71 × 10¹¹ ± 0.0071 n/cm² 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 × 10¹¹ ± 0.0042 n/cm² s, 2.53 × 10¹¹ ± 0.0074 n/cm² s and 4.73 × 10¹⁰ ± 0.0162 n/cm² s for the thermal, epithermal and fast neutron flux, respectively. The epicadmium design recorded a peak thermal flux of 7.08 × 10¹¹ ± 0.0033 n/cm² s and an epithermal flux of 2.09 × 10¹¹ ± 0.006 n/cm² 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 × 10¹¹ ± 0.0079 n/cm² 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.     

Target optimization for a 10-MeV E-beam neutron converter

Accelerator-based target design and optimization is an approach for neutron generation. The target plays an important role for a neutron source on an electron accelerator. For optimizing a neutron source using 10 MeV electron beams of Rhodotron-TT200, Pb, Ta, or W alloys with Be were calculated as photo-neutron converter. The neutron yield, flux and energy were simulated using the MCNPX code. The results indicate that a 10 MeV electron beam is capable of producing high-intensity neutron flux of 1013n·cm–2·s–1 with average energy of 0.8 MeV.     

Thermal hydraulic effect of fuel plate surface roughness

This study presents surface roughness measurements characteristic of the pre-film layer applied to a typical Advanced Test Reactor (ATR) fuel plate. This data is used to estimate the friction factor for thermal hydraulic flow calculations of a Gas Test Loop (GTL) system proposed for incorporation into ATR to provide a fast neutron flux environment for the testing of nuclear fuels and materials. To attain the required neutron flux, the design includes booster fuel plates clad with the same aluminum alloy as the ATR driver fuel and cooled with water supplied by the ATR primary coolant pumps. The objectives of this study are to (1) determine the surface roughness of the protective boehmite layer applied to the ATR driver fuel prior to reactor operations in order to specify the machining tolerances for the surface finish on simulated booster fuel plates in a GTL hydraulic flow test model and (2) assess the consequent thermal hydraulic impact due to surface roughness on the coolability of the booster fuel with a similar pre-film layer applied. While the maximum roughness of this coating is specified to be 1.6 μm (63 μin.), no precise data on the actual roughness were available. A representative sample coupon autoclaved with the ATR driver fuel to produce the pre-film coating was analyzed using optical profilometry. Measurements yielded a mean surface roughness of 0.53 μm (21 μin.). Results from a sensitivity study show that a ±15% deviation from the mean measured surface finish would have a minimal effect on coolant temperature, coolant flow rate, and fuel temperature. However, frictional losses from roughnesses greater than 1.5 μm (∼60 μin.) produce a marked decrease in flow rate, causing fuel and coolant temperatures to rise sharply.     

Sensitivity of fuel failure detection by the delayed neutron measurement in the primary cooling circuit of HANARO

The sensitivity of the fuel failure detection system based on the delayed neutron measurement in the primary cooling circuit of a research reactor, HANARO is investigated. The neutrons around the primary cooling pipe during normal operation of HANARO are measured with BF₃ detector, and their count rate is 900 cps. They are regarded as photoneutrons due to the high energy gamma-rays from N-16 and delayed neutrons from the fission of the uranium contaminated on the fuel surface. The contribution of each neutron source is analyzed by measuring the changes of the neutron counts before and after the abrupt shutdown of reactor. In order to estimate the sensitivity of the fuel failure detection, the neutron count rate of BF₃ detector is predicted by Monte Carlo calculation. The generation, transportation and detection of the photoneutrons and the delayed neutrons are simulated for the geometry similar to the experiments. From the calculations and experiments, it is ascertained that the photoneutron contribution to the total count rate is about 20–30%, and that the delayed neutron count rate is expected to about 720 cps. The fission rate in the flow tube of the reactor core by the surface contamination is obtained from the deduced delayed neutron count rate, and it is estimated to 1.66 × 10⁵ fissions/cm³ s. From the MCNP calculation, it is confirmed that this fission rate can originate from the contaminated uranium of 120 μg, which is about 13% of the maximum allowable surface contamination on the fuel surface. The sensitivity of U-235 mass detection by the delayed neutron measurement can be concluded to about 0.2 μg-U²³⁵/cps. Thus, it is confirmed that the delayed neutron detection is sensitive enough to monitor the fuel failure, and that the neutron count rate is high enough for stable signal with short counting time.     

Measurement of thermal neutron and resonance integral cross sections of the reaction V(n,γ)V using a 20 Ci Am–Be isotopic neutron source

The thermal neutron capture cross section (σ_o) and the resonance integral (I_o) of the ⁵¹V(n,γ)⁵²V reaction were measured with an activation method to provide fundamental data for reactor calculation, activation analysis, and other theoretical and experimental uses concerning the interaction of neutron with matter. The vanadium and manganese samples were irradiated within and without a Cd shield case using a 20 Ci Am–Be neutron source. The activities of the samples were measured using gamma-ray spectroscopy. The thermal neutron capture cross section and the resonance integral were determined relative to the reference reaction ⁵⁵Mn(n,γ)⁵⁶Mn and the values obtained are 5.16 ± 0.19 barns and 2.53 ± 0.1 barns respectively. The previous measurements of the σ_o and I_o of the reaction ⁵¹V(n,γ)⁵²V were reviewed and the difference between the present values and the previous results were discussed.     

Investigative studies on effect of reflector thickness on the performance of low enriched uranium-fueled miniature neutron source reactors

Neutronics analyses were performed on the 30 kW(th) GHARR-1 facility to investigate the effects on increased beryllium annular reflector thickness on nuclear criticality safety and on the neutron flux levels in the experimental channels. The investigative study was carried out using the Monte Carlo code MCNP on a hypothetical LEU UO₂ core theoretically enriched to 12.6% and having the same core configuration as the present 90.2% enriched HEU U-Al core. The analyses were performed on four models consisting of a reference model with 10.2 cm annular reflector thickness and three new design modification models with increased reflector thickness of 10.3, 10.4 and 10.5 cm respectively. The simulations indicated average thermal neutron fluxes of (9.80 ± 0.0017)E+11 n/cm² s in the inner irradiation channels for the reference model, indicating a 2% decrease with respect to the nominal flux of 1.00E+12 n/cm² s. Relatively lower neutron fluxes were obtained for the modification models with an average of (9.79 ± 0.0017)E+11 n/cm² s, representing losses of 2.01% and 0.01% with respect to the HEU core and reference LEU model.     

Thermal neutron cross section determination of short-to-medium lived nuclides using a 20 Ci Am–Be neutron source

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 ¹²⁷I(n,γ)¹²⁸I, ¹⁵²Sm(n,γ)¹⁵³Sm,¹⁵⁴Sm(n,γ)¹⁵⁵Sm, and ²³⁸U(n,γ)²³⁹U were determined by the method of foil activation using ⁵⁵Mn(n,γ)⁵⁶Mn as a reference reaction. The experimental samples with and without a Cd cover were irradiated in an isotropic neutron field of a 20 Ci ²⁴¹Am–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 ¹²⁷I(n,γ)¹²⁸I, ¹⁵²Sm(n,γ)¹⁵³Sm, ¹⁵⁴Sm(n,γ)¹⁵⁵Sm, and ²³⁸U(n,γ)²³⁹U were deduced from the analysis of obtained gamma-ray spectra. The thermal neutron capture cross section values for ¹²⁷I(n,γ)¹²⁸I, ¹⁵²Sm(n,γ)¹⁵³Sm, ¹⁵⁴Sm(n,γ)¹⁵⁵Sm, and ²³⁸U(n,γ)²³⁹U 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.     

Activation cross sections of deuteron-induced nuclear reactions on hafnium

Experimental excitation functions for deuteron-induced reactions up to 20 MeV on high purity natural hafnium were measured with the activation method using a stacked foil irradiation technique. Metallic hafnium foils with thickness of 10 μm were stacked together with 50 μm thick aluminium and 12 μm thick titanium foils. The aluminium foils served as energy absorber while the titanium foils were used to monitor the energy and intensity of the bombarding deuteron beam. From a detailed remeasurement of the complete excitation function of the ^natTi(d,x)⁴⁸V monitor reaction it was possible to adopt the proper incident energy and beam intensity by comparing the results with the recommended values. High resolution off-line gamma-ray spectrometry was applied to assess the activity of each foil. From the measured activity independent and/or cumulative elemental or isotopic cross section data for production of Ta, Hf and Lu radioisotopes by (d,x) reactions were determined. No experimental cross section data have been published earlier for these reactions in the investigated energy region. The work focuses on the production of ^177gLu that one of the promising radionuclides for small tumor therapy due to appropriate average energy of the emitted β-particles and the main gamma-rays that are suitable for detection by gamma-camera.The presented experimental data and results predicted by the TALYS theoretical code are compared. Thick target yields for production of the investigated radionuclides were calculated from a fit to our experimental excitation curves.     

Measurement of the total neutron cross-section and resonance parameters of molybdenum using pulsed neutrons generated by an electron linac

The neutron total cross-sections of molybdenum were measured in the neutron energy region from 0.01 eV to 200 eV by using the time-of-flight method at the Pohang Neutron Facility, which consists of an electron linear accelerator, a water-cooled tantalum target with a water moderator, and a 12-m long time-of-flight path. A ⁶Li-ZnS(Ag) scintillator with a diameter of 12.5 cm and a thickness of 1.6 cm was used as a neutron detector, and a high purity natural molybdenum metallic disc with a diameter of 6.2 cm and a thickness of 3 mm thickness was used for the neutron transmission measurement. Notch filters composed of Co, In, Cd were used to estimate the background level and to calculate the length of neutron flight path. In order to reduce the gamma-ray background from Bremsstrahlung and from neutron capture, we employed a neutron-gamma separation system based on their different pulse shapes. The present measurement was compared with the existing experimental and the evaluated data. The resonance parameters of Mo isotopes were extracted from the transmission by using the SAMMY code and were compared with other previous reported results.