Monte Carlo simulation of a conceptual thermal column in the Syrian miniature neutron source reactor using MCNP-4C

Based on probabilistic approach, the MCNP-4C code has been used effectively to simulate the Syrian MNSR reactor core and all its surrounding components in three dimensions, including a preliminary conceptual design of a thermal column to be installed later. For verification and validation purposes, reactor calculations include: criticality and control rod worth. Values of these parameters are 1.00517 and 6.54 mk, respectively. The thermal column is to be installed in the water of the reactor pool. Optimal conditions for this thermal column were tested using the already developed model. Optimization focused on the most suitable position for placement of the column in the water pool, dimensions, and material. The aim was to have a thermal neutron flux of 1 × 10⁹ n cm⁻² s⁻¹ in the center of thermal column, and resonant and fast neutron fluxes to be as low as possible as well.     

A D-D neutron generator using a titanium drive-in target

A D-D neutron generator was developed with an intensity of 10⁸ n/s. A helicon plasma ion source was used to produce a large current deuteron beam, and neutrons were generated by irradiating the deuteron beam on a titanium drive-in target made of commercial pure titanium. The neutron generator was test-run for several hundred hours, and the performances were investigated. The available range of the deuteron beam current was 0.8-8 mA and the beam could be accelerated up to 97.5 keV. The maximum neutron generation rate in the test-runs was 1.9 × 10⁸ n/s, which was achieved by irradiating a 7.6 mA deuteron beam at 94.0 keV on a 0.5 mm-thick target. The operation of the neutron generator was fairly stable, such that the neutron generation rate was not altered by high voltage breakdowns during the test-runs. Neutron generation efficiency was rated as low as 10% when compared to an ideal case of irradiating a 100% monatomic deuteron beam on a perfect TiD₂ target. Factors causing the low efficiency were suggested and discussed.     

On the impact of neutron beam divergence and scattering on the quality of transmission acquired tomographic images

The impact of the divergence of a thermal neutron beam and the scattered neutrons on the quality of tomographic images acquired by transmission have been evaluated by using a third generation tomographic system incorporating neutron collimators under several different arrangements. The system equipped with a gaseous position sensitive detector has been placed at the main channel outlet of the Argonauta Research Reactor in Instituto de Engenharia Nuclear (CNEN-Brazil) which furnishes a thermal neutron flux of 2.3 × 10⁵ n cm⁻² s⁻¹. Experiments have then been conducted using test-objects with well-known inner structure and composition to assess the influence of the collimators arrangement on the quality of the acquired images. Both, beam divergence and scattering - expected to spoil the image quality - have been reduced by using properly positioned collimators between the neutron source and the object, and in the gap between the object and the detector, respectively. The shadow cast by this last collimator on the projections used to reconstruct the tomographic images has been eliminated by a proper software specifically written for this purpose. Improvement of the tomographic images has been observed, demonstrating the effectiveness of the proposed approach to improve their quality by using properly positioned collimators.     

Design calculation of an epithermal neutronic beam for BNCT at the Syrian MNSR using the MCNP4C code

This article describes the design calculation of an epithermal neutronic beam for the boron neutron capture therapy at the Syrian MNSR by using the MCNP4C code and ENDF/B-V cross-section library. To produce a high flux of epithermal neutrons at the beam exit, the moderator/filter from Al, Cd, Fluental and Bi was used with Pb as reflector for neutrons along the beam. In addition, the Bi lined collimator with Li₂CO₃-PE and Pb at the end. The calculated beam parameters under 30.0 kW of reactor power at the beam exit are Ф_epi = 2.83 × 10⁸ n/cm² s, D_f/Ф_epi = 7.98 × 10⁻¹¹ cGy cm²/n, D_γ/Ф_epi = 1.70 × 10⁻¹¹ cGy cm²/n, Φ_epi/Φ_the = 0.05 and J_n⁺/Ф_n = 0.77. As well as, the calculated values of the advantage depth and advantage ratio are 7.51 cm and 3.49, respectively. If such beam was built into the Syrian MNSR the scientific applications of the reactor would increase.     

Measurement of keV-neutron capture cross-sections and capture γ-ray spectra of Fe and Fe

The neutron capture cross-sections and the capture γ-ray spectra of ⁵⁶Fe and ⁵⁷Fe have been measured in the neutron energy range from 10 to 90 keV. Pulsed keV-neutrons were produced from the ⁷Li(p,n)⁷Be reaction by bombarding a lithium target with a 1.5-ns bunched proton beam from a 3 MV Pelletron accelerator. The incident neutron spectrum on the capture sample was measured using a time-of-flight method with a ⁶Li-glass detector. The capture γ-rays emitted from an iron or standard gold sample were detected with a large anti-Compton NaI(Tl) spectrometer. The capture yield of the iron or gold sample was obtained by applying a pulse-height weighting technique to the corresponding capture γ-ray pulse-height spectrum. The capture cross-sections of ^56,57Fe were derived with errors less than 5% using the standard capture cross-sections of ¹⁹⁷Au. The capture γ-ray spectra were obtained by unfolding the observed capture γ-ray pulse-height spectra. The present results for the capture cross-sections were compared with the previous measurements and the evaluated values of ENDF/B-VII.0 and JENDL-3.3. The Maxwellian-averaged capture cross-sections of ⁵⁶Fe and ⁵⁷Fe at 30 keV are derived as 12.22 ± 2.06 mb and 44.48 ± 7.56 mb, respectively.     

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.     

CR-39 detector based thermal neutron flux measurements, in the photo neutron project

PhoNeS (photo neutron source) is a project aimed at the production and moderation of neutrons by exploiting high energy linear accelerators, currently used in radiotherapy. A feasibility study has been carried out with the scope in mind to use the high energy photon beams from these accelerators for the production of neutrons suitable for boron neutron capture therapy (BNCT). Within these investigations, it was necessary to carry out preliminary measurements of the thermal neutron component of neutron spectra, produced by the photo-conversion of X-ray radiotherapy beams supplied by three LinAcs: 15 MV, 18 MV and 23 MV. To this end, a simple passive thermal neutron detector has been used which consists of a CR-39 track detector facing a new type of boron-loaded radiator. Once calibrated, this passive detector has been used for the measurement of both the thermal neutron component and the cadmium ratio of different neutron spectra. In addition, bubble detectors with a response highly sensitive to thermal neutrons have also been used. Both thermal neutron detectors are simple to use, very compact and totally insensitive to low-ionizing radiation such as electrons and X-rays. The resultant thermal neutron flux was above 10⁶ n/cm²s and the cadmium ratio was no greater than 15 for the first attempt of photo-conversion of X-ray radiotherapy beams.     

Calculation of fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor using the GETERA code

Calculations of the fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burntup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission products and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer, has a bigger library of isotopes, and is more accurate.     

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.     

Measurement of thermal neutron cross section and resonance integral for (n,γ) reaction in Sm

This study implies that ⁵⁵Mn(n,γ)⁵⁵Mn monitor reaction may be a convenient alternative comparator for the activation method and thus, it was used for the determination of thermal neutron cross section (TNX) and the resonance integral (RI) of the reaction ¹⁵²Sm(n,γ)¹⁵³Sm. The samples of MnO₂ and Sm₂O₃ diluted with Al₂O₃ powder were irradiated within and without a cylindrical 1 mm-Cd shield case in an isotropic neutron field obtained from the ²⁴¹Am–Be neutron sources. The γ-ray spectra from the irradiated samples were measured by high resolution γ-ray spectrometry with a calibrated n-type Ge detector. The correction factors for γ-ray attenuation, thermal neutron and resonance neutron self-shielding effects and epithermal neutron spectrum shape factor (α) were taken into account in the determinations. The thermal neutron cross section for ¹⁵²Sm(n,γ)¹⁵³Sm reaction has been determined to be 204.8 ± 7.9 b at 0.025 eV. This result has been obtained relative to the reference thermal neutron cross section value of 13.3 ± 0.1 b for the ⁵⁵Mn(n,γ)⁵⁶Mn reaction. For the TNX, most of the experimental data and evaluated one in JEFF-3.1, ENDF/B-VI, JENDL 3.3 and BROND 2.0, in general, agree well with the present result. The RI value for ¹⁵²Sm(n,γ)¹⁵³Sm reaction has also been determined to be 3038 ± 214 b, relative to the reference value of 14.0 ± 0.3 b for the ⁵⁵Mn(n,γ)⁵⁶Mn monitor reaction, using a 1/E^1+α epithermal neutron spectrum and assuming Cd cut-off energy of 0.55 eV. In surveying literature, the existing experimental and evaluated data for the RI values are distributed from 1715 to 3462 b. However, when the Cd cut-off energy is defined as 0.55 eV, the present RI value agrees with some previously reported RI values, 3020 ± 163 b by Simonits et al., 3141 ± 157 by Van Der Linden et al., and 2962 ± 54 b by Kafala et al., within the limits of error.     

Time dependent burn-up and fission products inventory calculations in the discharged fuel of the Syrian MNSR

The Syrian Miniature Neutron Source Reactor (MNSR), a 30 kW, 89.8% HEU fueled (U-Al), went critical in March, 1996. By operating the reactor at nominal power for 2.5 h/day, the estimated core life is 10 years. This paper presents the results of fuel burn-up and depletion analysis of the MNSR fuel lattice using the ORIGEN 2 code. A one-group cross-section data base for the ORIGEN 2 computer code was developed for the Syrian MNSR research reactor. The ORIGEN 2 predicted burn-up dependent actinide compositions of MNSR spent fuel using the newly developed data base show a good agreement with the published results in the literature. In addition, the burn-up characteristics of MNSR spent fuel was analyzed with the new data base. Finally, to study the effect of burn-up on the reactivity, the microscopic cross-sections of the fission products calculated by the WlMS code (using the number densities of fission products generated by the ORIGEN 2 code as a function of burn-up time), were used as an input for the CITATION code calculations. The results contained in this paper could be used in performing criticality safety analysis and shielding calculations for the design of a spent fuel storage cask for the MNSR core.     

Radiation-induced stress relaxation in high temperature water of type 316L stainless steel evaluated by neutron diffraction

Weld beads on plate specimens made of type 316L stainless steel were neutron-irradiated up to about 2.5 × 10²⁵ n/m² (E > 1 MeV) at 561 K in the Japan Material Testing Reactor (JMTR). Residual stresses of the specimens were measured by the neutron diffraction method, and the radiation-induced stress relaxation was evaluated. The values of σ_x residual stress (transverse to the weld bead) and σ_y residual stress (longitudinal to the weld bead) decreased with increasing neutron dose. The tendency of the stress relaxation was almost the same as previously published data, which were obtained for type 304 stainless steel. From this result, it was considered that there was no steel type dependence on radiation-induced stress relaxation. The neutron irradiation dose dependence of the stress relaxation was examined using an equation derived from the irradiation creep equation. The coefficient of the stress relaxation equation was obtained, and the value was 1.4 (×10⁻⁶/MPa/dpa). This value was smaller than that of nickel alloy.     

MgO single-crystal as an efficient thermal neutron filter

A formula is given which, for neutron energies in the range 10⁻⁴ < E < 10 eV, permits calculation of the nuclear capture, thermal diffuse and Bragg scattering cross-sections as a function of magnesium oxide (MgO) temperature and crystal parameters. Computer program has been developed to calculate the total neutron cross-section and transmission through mono-crystalline MgO. The calculated neutron transmission and effective attenuation coefficient values for MgO-single crystal at different temperatures are compared with measured ones. An overall agreement is indicated between the formula fits and experimental data. A feasibility study for the use of MgO-single crystal is discussed in terms of the optimum MgO-single crystal thickness, mosaic spread, temperature and cutting plane for efficient transmission of thermal-reactor neutrons, and also for rejection of the accompanying fast neutrons and gamma rays.     

Properties of Na implanted targets

Ten types of ²³Na implanted targets have been fabricated for the purposes of investigating the effects of proton beam bombardment on the implanted sodium distribution. Targets were implanted at energies of E_Na = 10-30 keV using copper, tantalum, and nickel as host materials. Thin layers (100-200 Å) of chromium and gold were also evaporated over some of the targets to provide a protective layer for the implanted sodium. The ²³Na(p, γ) resonance at a lab proton energy of E_p = 309 keV was used to determine the implanted distribution. Successive resonance profile measurements are presented for each implanted target, and the concurrent loss of ²³Na resulting from beam bombardment is reported. The calculated temperature rise of the targets indicates that beam heating has a negligible effect on the implanted sodium distribution, and that the principal mechanism for ²³Na loss during beam bombardment is sputtering.     

Estimation of neutron production from accelerator head assembly of 15 MV medical LINAC using FLUKA simulations

For the production of a clinical 15 MeV photon beam, the design of accelerator head assembly has been optimized using Monte Carlo based FLUKA code. The accelerator head assembly consists of e-γ target, flattening filter, primary collimator and an adjustable rectangular secondary collimator. The accelerators used for radiation therapy generate continuous energy gamma rays called Bremsstrahlung (BR) by impinging high energy electrons on high Z materials. The electron accelerators operating above 10 MeV can result in the production of neutrons, mainly due to photo nuclear reaction (γ, n) induced by high energy photons in the accelerator head materials. These neutrons contaminate the therapeutic beam and give a non-negligible contribution to patient dose. The gamma dose and neutron dose equivalent at the patient plane (SSD = 100 cm) were obtained at different field sizes of 0 × 0, 10 × 10, 20 × 20, 30 × 30 and 40 × 40 cm², respectively. The maximum neutron dose equivalent is observed near the central axis of 30 × 30 cm² field size. This is 0.71% of the central axis photon dose rate of 0.34 Gy/min at 1 μA electron beam current.     

Measurements of neutron total cross-sections and resonance parameters of erbium at the Pohang Neutron Facility

We measured neutron total cross-sections of natural erbium in the neutron energy region from 0.2 to 120 eV by using the neutron 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 group of high-purity natural erbium metallic plates with various thickness was used for the neutron transmission measurements. The present measurement was compared with the existing experimental and the evaluated data. The resonance parameters of ¹⁶⁶Er, ¹⁶⁷Er, ¹⁶⁸Er, and ¹⁷⁰Er in the neutron energy region below 120 eV were extracted from the transmission by using the multilevel R-matrix SAMMY code and were compared with the evaluated data from ENDF/B VII.0 and other previous reported results.     

An innovative research reactor design

A new and innovative core design for a research reactor is presented. It is shown that while using the standard, low enriched uranium as fuel, the maximum thermal flux per MW of power for the core design suggested and analyzed here is greater than those found in existing state of the art facilities without detrimentally affecting the other design specs. A design optimization is also carried out to achieve the following characteristics of a pool type research reactor of 10 MW power: high thermal neutron fluxes; sufficient space to locate facilities in the reflector; and an acceptable life cycle. In addition, the design is limited to standard fuel material of low enriched uranium. More specifically, the goal is to maximize the maximum thermal flux to power ratio in a moderate power reactor design maintaining, or even enhancing, other design aspects that are desired in a modern state of the art multi-purpose facility. The multi-purpose reactor design should allow most of the applications generally carried out in existing multi-purpose research reactors. Starting from the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, an azimuthally asymmetric cylindrical core design with an inner and outer reflector, is developed. More specifically, one half of the annular core (0 < θ < π) is thicker than the other half. Two variations of the design are analyzed using MCNP, ORIGEN2 and MONTEBURNS codes. Both lead to a high thermal flux zone, a moderate thermal flux zone, and a low thermal flux zone in the outer reflector. Moreover, it is shown that the inner reflector is suitable for fast flux irradiation positions. The first design leads to a life cycle of 41 days and high, moderate and low (non-perturbed) thermal neutron fluxes of 4.2 × 10¹⁴ n cm⁻² s⁻¹, 3.0 × 10¹⁴ n cm⁻² s⁻¹, and 2.0 × 10¹⁴ n cm⁻² s⁻¹, respectively. Heat deposition in the cladding, coolant and fuel material is also calculated to determine coolant flow rate, coolant outlet temperature and maximum fuel temperature under steady-state operating conditions. Finally, a more compact version of the asymmetric core is developed where a maximum (non-perturbed) thermal flux of 5.0 × 10¹⁴ n cm⁻² s⁻¹ is achieved. The core life of this more compact version is estimated to be about 23 days.     

Fluence and dose measurements for an accelerator neutron beam

The 3 MV Van de Graaff accelerator at McMaster University accelerator laboratory is extended to a neutron irradiation facility for low-dose bystander effects research. A long counter and an Anderson-Braun type neutron monitor have been used as monitors for the determination of the total fluence. Activation foils were used to determine the thermal neutron fluence rate (around 10⁶ neutrons s⁻¹). Meanwhile, the interactions of neutrons with the monitors have been simulated using a Monte Carlo N Particle (MCNP) code. Bystander effects, i.e. damage occurring in cells that were not traversed by radiation but were in the same radiation environment, have been well observed following both alpha and gamma irradiation of many cell lines. Since neutron radiation involves mixed field (including gamma and neutron radiations), we need to differentiate the doses for the bystander effects from the two radiations. A tissue equivalent proportional counter (TEPC) filled with propane based tissue equivalent gas simulating a 2 μm diameter tissue sphere has been investigated to estimate the neutron and gamma absorbed doses. A photon dose contamination of the neutron beam is less than 3%. The axial dose distribution follows the inverse square law and lateral and vertical dose distributions are relatively uniform over the irradiation area required by the biological study.     

Measurement of neutron flux distribution in the irradiation channel in the Ghana Research Reactor-1 using Monte Carlo method

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 ± 0.0007. Radially, the thermal, epithermal and fast neutron flux in the inner irradiation channel range from 1.15 × 10¹² n/cm².s ± 0.1018 × 10¹¹ − 1.19 × 10¹² n/cm².s ± 0.1172 × 10¹¹, 1.21 × 10¹² n/cm².s ± 0.1014 × 10¹¹ − 1.36 × 10¹² n/cm².s ± 0.1038 × 10¹¹ and 2.47 × 10¹¹ n/cm².s ± 0.1120 × 10¹⁰ − 2.97 × 10¹¹ n/cm².s ± 0.1255 × 10¹⁰ respectively. For the outer channel, the flux range from 7.14 × 10¹¹ n/cm².s ± 0.1381 × 10¹⁰ − 7.38 × 10¹¹ n/cm².s ± 0.208 × 10¹⁰ for thermal, 1.94 × 10¹¹ n/cm².s ± 0.1014 × 10¹⁰ − 2.51 × 10¹¹ n/cm².s ± 0.1281 × 10¹⁰ for epithermal and 3.69 × 10¹⁰ n/cm².s ± 0.8912 × 10⁸ − 5.14 × 10¹⁰ n/cm².s ± 0.1009 × 10⁹ for fast. The results have shown that there are flux variations within the irradiation channels both axially and radially.     

Neutron irradiation-induced dimensional changes in MEMS glass substrates

A study is made of radiation-induced expansion/compaction in Pyrex^® (Corning 7740) and Hoya SD-2^® glasses, which are used as substrates for MEMS devices. Glass samples were irradiated with a neutron fluence composed primarily of thermal neutrons, and a flotation technique was employed to measure the resulting density changes in the glass. Transport of Ions in Matter (TRIM) calculations were performed to relate fast (∼1 MeV) neutron atomic displacement damage to that of boron thermal neutron capture events, and measured density changes in the glass samples were thus proportionally attributed to thermal and fast neutron fluences. Pyrex was shown to compact at a rate of (in Δρ/ρ per n/cm²) 8.14 × 10⁻²⁰ (thermal) and 1.79 × 10⁻²⁰ (fast). The corresponding results for Hoya SD-2 were 2.21 × 10⁻²¹ and 1.71 × 10⁻²¹, respectively. On a displacement per atom (dpa) basis, the compaction of the Pyrex was an order of magnitude greater than that of the Hoya SD-2. Our results are the first reported measurement of irridiation-induced densification in Hoya SD-2. The compaction of Pyrex agreed with a previous study. Hoya SD-2 is of considerable importance to MEMS, owing to its close thermal expansivity match to silicon from 25 to 500°C.