Miniature Neutron Source Reactor Research Articles

A computational- experimental model of reactor kinetic for investigating the linearity response of in-core neutron detectors of a low power research reactor

In this study, the linear response of fission chamber (FC) neutron detectors in Isfahan miniature neutron source reactor (MNSR) is investigated through experiments and calculations at different powers of the reactor. The linearity of the response of the main neutron detector in the MNSR is important due to its safety functions in reactor power control. Additionally, determining the linear response region of this detector presents a linear neutron dosimetry field even outside of the reactor core, i.e. at the outlet of external beam tubes. Moreover, it is very difficult to remove and calibrate this neutron detector that has been installed in the reactor core for a long time. For this purpose, an alternative method for measuring the linearity of the response of a neutron detector is extended by investigating the kinetic behavior of reactor. The aim of this method is to calculate the delayed neutrons population in the reactor core and compare with the FC detector response. In this method, the numerical solution method is used to solve the point kinetic equation with 15 delayed neutron groups (6 delayed neutron groups and 9 photo-neutron groups) after inserting a large negative reactivity into the MNSR core by considering temperature and xenon reactivity feedbacks during a coupling procedure. To obtain the xenon reactivity feedback, xenon concentration is calculated by coupling the 135Xe and 135I production equations and the εpPFnlPTnl parameter is obtained by an experimental method. Furthermore, the value of fission cross section (Σf) is calculated using a validated model in the Dragon code. The large reactivities are inserted at two initial power of 150 W and 30 kW to investigate the response of the main FC neutron detector in a cold xenon-free condition and by considering temperature and xenon reactivity feedbacks, respectively. The results demonstrate that the response of the MNSR FC detector is linear in the neutron flux range between 5 × 107 to 1 × 1012 n.cm−2.s−1, which is equivalent to reactor power of 1.5 W to 30 kW. This innovative method can be applied to calibrate the neutron detectors in other research and power reactors.

Experimental evaluation of transition rate of sapphire crystal for thermal and fast neutrons using MNSR vertical neutron beam line

Using a perfect single crystal as a neutron filter allows us to have a thermal neutron beam with almost no background of fast neutrons. Single crystals of Al2O3 (sapphire) have proven to be effective filters for fast neutrons and are incorporated into neutron instruments. The present work would experimentally investigate c-axis neutron transmission rate by using different crystal thicknesses. In fact, the optimal thickness for sapphire filter is the one that maximizes the transmission of low energy neutrons and minimizes the transmission of fast neutrons, if there is no significant decrease in thermal neutron flux. In addition, neutron-filtering power of a-axis and c-axis sapphire crystals were compared with each other using different tests on a 2.5 cm slab of the sapphire crystals. The experimental tests were carried out by means of the available neutron flux top of the vertical neutron beam line of the Isfahan Miniature Neutron Source Reactor (MNSR) in two methods of foil activation and flux monitoring. In addition, the thermal and fast neutron dose rate reduction was discussed by using different thicknesses of the c-axis crystal.

Feasibility study for the adoption of multi-capsule irradiation protocol in the conduct of k0-based INAA using the GHARR-1 MNSR

The Ghana Research Reactor-1 (GHARR-1) is currently a 23 cm length LEU core Miniature Neutron Source Reactor (MNSR) with a 13 % U-235 enrichment having 335 fuel rods, 15 dummy rods and a central control rod for neutron regulation. It has 10 accessible irradiation channels of approximate length of 17 cm and formed part of the shim tray structure which was considered for active routine experiment in the present work. Unlike the routine characterization of the inner irradiation channel of the Ghana Research Reactor- 1 facility which involves mainly the introduction of a single irradiation capsule of length, 5 cm loaded with samples, the present study adopted a multi-capsule scheme in which three (3) capsules each of length 5 cm were introduced into the 23 cm long irradiation channel to scientifically interrogate the feasibility and hence provide a sound basis for extending and improving the accessible irradiation space by 60% during the utilization of the GHARR-1, especially for irradiation involving intermediate and long lived radionuclides. The objective is also to achieve the characterization of neutron spectrum and determine their spatial distribution for close to the full length of the irradiation channel. Validation protocols based on k0 method were developed through the analysis of some reference materials for a careful study of irradiation, decay and counting scheme which achieved optimum radionuclide selectivity. The overall approach adopted for the flux characterization involves the preparation of flux monitors, packaging of three (3) capsules each of both bare and cadmium cover samples, irradiation at bottom, middle and top spatial demarcation of the irradiation channel, with each demarcation being the region of the 5 cm length of the bottom, middle and top irradiation capsules. Sample counting was undertaken using the HPGe detector after the samples were allowed specific decay time after irradiation and prior to counting. Flux monitors were used for the flux characterization and reference materials were used for the validation protocol. Spectrum acquisition was made possible through the use of the Gamma Vision Software and spectrum parameters (thermal to epithermal neutron flux ration (f), epithermal neutron shaping factor (α), thermal, epithermal, and fast fluxes) were determined. Results obtained showed increasing f-value across the irradiation column as, 18.5 ± 1.7, 21.0 ± 2.2 and 23.0 ± 7.08 respectively from the bottom capsule to the top capsules. The corresponding epithermal neutron shaping factor (α-value) varied as, −0.096 ± 0.029, −0.18 ± 0.036 and −0.20 ± 0.06 from the bottom to the top capsule. The experimental results determined in the bottom, middle and top capsule irradiation column for thermal, epithermal and fast fluxes are 4.60 × 1011±2.5 × 1010, 2.49 × 1010±5.98 × 108, 9.24 × 1010±2.2 × 109; 4.21 × 1011±1.01 × 1010, 2.01 × 1010±4.82 × 108, 4.81 × 1010±1.15 × 109; and 3.90 × 1011±9.36 × 109, 1.65 × 1010±3.90 × 108, 4.82 × 1010±1.16 × 109 respectively. The validation protocol using standard reference materials and treating each capsule with separate reactor characterization parameters indicated respective z-score distribution within a 95% confidence interval.

Development and characterization of fission chamber neutron detectors in Isfahan miniature neutron source reactor

In this research, two fission chamber neutron detectors have been designed and constructed, and their performance has been evaluated in the radiation field of the Isfahan miniature reactor neutron beam line. Each detector consists of two coaxial cylinders called anode and cathode made of stainless steel. The outer layer of the anode is covered with a very thin layer of uranium-235 oxide as fissile material, and a mixture of argon and nitrogen gas is used as the filling gas. The thickness of the deposited layer on the anode is different, so that on one of them, named FC400A, the thickness of the uranium oxide layer was more than 5 μm, and on the other one, named FC400B, it was less than 1.5 μm. A spectrum collection software has also been designed and developed in LabVIEW. The developed detectors are irradiated in the neutron field resulting from the neutron radiography beamline of the miniature neutron source reactor in the range of neutron flux of about 2E+02 to 2E+05 n.cm−2.s−1. In order to determine the sensitivity of the detection response, the results of these two detectors are also compared with a commercial fission chamber neutron detector. The results showed that the working voltage range of these detectors is in the range of 400–600 V and both of these detectors have a suitable linear response in the mentioned neutron flux range. Also, the sensitivity of FC400A and FC400B detectors is 0.1 and 1 counts/1E+04 neutrons, respectively.

Determination of the linear behavior of FC detectors in Isfahan MNSR using ex-core offline and online experiments

The linear behavior of fission chamber (FC) neutron detectors in Isfahan miniature neutron source reactor (MNSR) was investigated experimentally. FC neutron detectors are installed near the reactor core to measure the produced neutron flux in the core. The linearity of the response of these detectors is important because the measured neutron flux by them is one of the important methods in determining and controlling the reactor power and therefore affects the safety of the reactor. For this purpose, the measured neutron fluxes by these detectors at different reactor powers were compared with the measured neutron flux at the output of the external beam tubes by different methods. In this study, two online and offline neutron flux detection methods were used to analysis the linear behavior of MNSR neutron detectors. The BF3 detector and solid-state nuclear track detector (SSNTD) were used to determine the neutron flux in the online and offline methods, respectively. The results of two diverse methods showed that the response of MNSR FC detectors for measuring the reactor power is linear with acceptable accuracy in the neutron flux range between 5×107 to 1012n.cm−2. s−1, which is equivalent to reactor power of 1.5 W to 30 kW.

Long-time irradiation effect on corrosion behavior of aluminum alloy in pool water of low-power research reactor

This study conducted an evaluation of the corrosion behavior of an aluminum alloy utilized in the Isfahan Miniature Neutron Source Reactor (MNSR). The component analyzed, dry channel (DC), had been exposed to radiation for 12 years in a water environment within the reactor pool since its installation. To determine the effect of radiation on the corrosion of the LT-21 aluminum alloy used in the DC, different parts of the pipe were sampled and various tests were performed. These tests included mechanical strengths (impact, and micro-hardening), XRD, TEM, SEM–EDS, and potentiodynamic polarization (PDP). The parameters measured included corrosion potential, corrosion rate, changes in microscopic structure, and mechanical properties of the aluminum alloy along the entire length of the DC. The neutron and gamma dose distribution along the height of the DC, which was 540 cm, was calculated to determine the correlation between the dose distribution and observed corrosion. The study found that the corrosion mechanisms were complex and resulted from the simultaneous presence of the DC in the pool water and radiation from the reactor core. The observed results are presented and discussed in this study.

Feasibility study of nanomaterials synthesis at MNSR research reactor through design and construction of a gamma irradiation cell

Herein, a gamma irradiation cell consisting of a borated polyethylene (BPE) cylinder with a cadmium inner wall was designed and constructed at Isfahan miniature neutron source reactor (MNSR) for synthesis of metal nanoparticles. The cell is placed in a dry channel irradiation site near the reactor core. The performance of this irradiation cell with appropriate shape and size was evaluated and the neutron flux (φ) and the neutron and gamma dose rates (Dn and Dg) inside it were calculated using MCNP code. The results showed that the designed cell could deliver an appropriate gamma dose rate to samples and reduce the thermal neutron flux from 5.82 × 1010 to 6.28 × 107n.cm−2.s−1, i.e. over 99.89% attenuation. Also, the gamma dose rate inside the designed neutron shield is calculated to be 3.83 kSv.h−1. To evaluate the designed and constructed cell, synthesis of Pd and Ag nanoparticles (NPs) on SiO2 supporter was done. Prepared nanocomposites were characterized by using XRD, FE-SEM, and EDXA techniques. Observed results showed metallic Pd and Ag NPs were produced successfully employing a designed gamma irradiation cell at the reactor via gamma-assisted route in nanometer sizes for the first time.

Design of an out-of-tank neutron beam line for prompt gamma neutron activation analysis at low-power Miniature Neutron Source Reactor

Research reactors are strong and stable neutrons sources, and various researches and applications were carried out using their neutron and gamma beamlines. Some research reactors, such as Miniature Neutron Source (MNSR) Reactors, do not have beam tube in their original design, which can be considered as a disadvantage. The design and installation of beamlines can develop their applications. In this research, the design and optimization of a suitable neutron beamline for PGNAA is done at Isfahan MNSR. Designs are done to reach the appropriate neutron beam parameters and taking into account the geometrical limitations of the MNSR. Neutron beam parameters including thermal neutron flux, thermal to total neutron flux, and thermal neutron flux to gamma dose ratio are considered. For this, vertical beamlines with different diameters are simulated and the optimal diameter is determined. Also, for optimization of beam parameters, a thermal column and gamma filter are considered and optimal conditions are determined.

Feasibility study on PGNAA experiments using a prototype neutron beam at Isfahan MNSR

In this study, the feasibility study of Prompt Gamma Neutron Activation Analysis (PGNAA) experiments at Isfahan Miniature Neutron Source Reactor (MNSR) has been done using a prototype neutron beam. This neutron beam was not originally created for this purpose, but previous studies on this neutron beam have shown that it has enough neutron flux to carry out some irradiations outside the reactor core. This prototype neutron beam provided a neutron flux of order 105 n·cm-2·s-1 and a circular diameter of 5 cm for the experiments. In addition, an appropriate setup for High Purity Germanium detector (HPGe) was done at the beam outlet, where the sample is irradiated. A simple shield was also installed to prevent radiation damage to the detector. Experiments were conducted for bulk samples of salt solution with different concentrations and the prompt gamma spectrums have been obtained and analyzed.

Design of an in-tank thermal neutron beam for PGNAA application at Isfahan MNSR

Calculations have been done in order to achieve a thermal neutron beam for Prompt Gamma Neutron Activation Analysis (PGNAA) applications using Miniature Neutron Source Reactor (MNSR) as neutron source. The MNSR is a 30 kW tank-in-pool reactor and there are ten small occupied in-tank irradiation positions. Also, there are four positions in order to install or remove aluminum Regulating Rod Tubes (RRTs) that are bigger than that ten irradiation positions. In this study, the eastern RRT has been selected for more simulations of thermal neutron beam. Thermal neutron flux (φth), the ratio of thermal neutron flux to gamma dose rate (φthDγ), and the ratio of thermal neutron flux to total neutron flux (φthφtot) have been considered in calculations as important neutron beam parameters. Some designs have been done and simulated using MCNP code. The effect of the height of beam inlet, the effect of the different materials and thickness of moderators, and the effect of different thickness and height of gamma filters on the beam parameters at the beam outlet have been investigated. The results of different designs showed that using appropriate heights and thicknesses, by using bismuth as gamma filter and heavy water as moderator, φth, φthφtot and φthDγ, are 1.64E + 06n.cm−2.s−1, 0.789, and 4.36E + 05n.cm−2.μSv−1 respectively. Also, by using bismuth as gamma filter and graphite as moderator, these parameters are 1.57E + 06n.cm−2.s−1, 0.761, and 4.69E + 05E + 05n.cm−2.μSv−1, respectively.

Application of SCALE code for reactor physics analysis of the as-built Nigeria miniature neutron source reactor operating with low enriched uranium fuel

This study presents a broad reactor physics analysis including, neutronic, burnup, and decay heat of the ‘as-built’ NIRR-1 LEU core performed using SCALE 6.2.3 code package. First, verification of neutronic parameters of the NIRR-1 was performed by modeling the reactor ‘as-built’ LEU core. Thereafter validation of the model was done by comparing calculated result with measured data during the commissioning of the reactor. The results showed that the estimated neutronic parameters using KENO-VI code of SCALE 6.2.3 code package are in close agreement with the measured data. Hence, KENO-VI can be used for detailed neutronic assessment of MNSRs. The total decay heat at the end of core life after operation for 216 EFPD was estimated to be 1620.15 W and was majorly due to the contribution from fission products which was estimated to be 1608.10 W. This value explains why MNSRs are cooled through natural convection of the coolant during operation and at shutdown conditions. Additionally, the burnup of the core was determined to be 0.53 wt% which is within the expected burnup for a typical MNSRs (less than1 wt%) with a core lifetime of about 200 EFPD (∼10 years). Furthermore, the mass of 239Pu buildup after 216 EFPD was far less than the IAEA category III nuclear material and indicates a huge advantage of research reactor conversion in reducing proliferation risk.

Development and characterization of a large thermal neutron beam for neutron radiography at Isfahan MNSR

The feasibility study of performing neutron radiography (NR) at the Isfahan Miniature Neutron Source Reactor (MNSR) was carried out in the previous work that was including calculations and experiments in support of the implementation of a prototype NR beam tube. Results were shown the obtained neutron beam from that prototype beam tube had appropriate conditions for NR and it is possible to obtain a large NR beam at this low power research reactor. The main disadvantage of the prototype beam tube was the small size of the beam port with diameter of 5 cm. In this study, a large diameter beam tube is designed and installed at the reactor. This beam tube has a telescopic divergent shape and is installed in the west side of the reactor pool. The neutron beam parameters and image quality of the obtained beam from this beam tube are measured based on ASTM-E545. Results showed that this neutron beam has good quality for NR experiments over a beam diameter of about 25 cm. Also, a simple and practical system, for fill and drain water to the beam tube, is designed and incorporated as a beam shutter in order to maintain the radiation safety.

Determination of Trace Elements in Different Pistachio Variety by Instrumental Neutron Activation Analysis Technique

Elemental analysis of nine samples of seven popular pistachios was performed using the neutron activation analysis technique. Thermal neutrons of Isfahan Miniature Neutron Source Reactor and related devices were used for activation analysis. The results show that the concentrations of calcium (Ca), copper (Cu), magnesium (Mg), and zinc (Zn) varied from 1670.20 to 1680.60, from 0.472 to 1.42, from 1170.50 to 1460.75, from 47.0 to 65.10 in mg/kg, respectively. Arsenic (As) was present in small amounts in some pistachio samples. Essential elements such as copper and zinc in Rafsanjan Akbari sample had higher amounts than other varieties.

Characterization of toothpastes for fluorine and other elements by INAA and ICP-OES

Abstract Fluoride toothpastes have made a significant contribution towards oral hygiene. This study proposes instrumental neutron activation analysis (INAA) as a suitable technique for determining fluorine in toothpastes. The method involved a 10 s irradiation, followed by 10–20 s decay, and counting for 10 s, using a pneumatic system and high purity germanium detector. All irradiations were performed at miniature neutron source reactor with thermal neutron flux 1 × 1012 cm−2 s−1. Since Na was present in all toothpastes, the quantification of fluorine required correction for 23Na(n,α)20F threshold reaction. Fluorine determination was validated by analysing the samples by ion chromatography. All toothpastes were also analysed for other elements by k 0-INAA and ICP-OES. Both techniques quantified 28 elements including Al, As, Ba, Ca, Ce, Cl, Co, Cr, Cu, F, Fe, Hf, K, La, Mg, Mn, Na, P, S, Sb, Sc, Si, Sm, Sn, Sr, Ta, Ti and Zn. This study, categorized toothpastes in five groups based on concentrations of Ca, P and Ti.

Assessment of neutronics performance of U3Si, U3Si2, and U–9Mo fuels for the support of MNSR conversion using SCALE 6.2.3 computational tools

In spite of the acceptability and production of Uranium Oxides (UO2) with Low Uranium Enrichment (LEU-UO2) fuel in commercial scale, the use of uranium–silicide (U–Si) in place of the UO2 is one of the promising concepts being proposed to increase the accident tolerance of nuclear fuels. In this study, SCALE 6.2.3 computational tools have been used to model the Nigerian Research Reactor-1 (NIRR-1) which is a typical Miniature Neutron Source Reactor (MNSR) and calculated the core neutronic parameters such as Clean Core Cold Excess Reactivity (CCER), Control Rod Worth (CRW), Shutdown Margin (SDM), Safety Reactivity Factor (SRF) and Neutron flux distribution for the Highly Enriched Uranium (HEU) and candidates LEU cores. The results showed a close agreement between the calculated and measured data. The bias between calculated and measured data was attributed to the computational limitation of this study that was caused by a limited number of particles histories used in criticality calculations. The neutronic performance of the U3Si2–Al, U3Si–Al and U9Mo–Al fuels indicated the suitability of these candidates LEU fuels for conversion of MNSRs from HEU to LEU. The observed reduction in the thermal flux of the candidate LEU fuels was attributed to higher inventory of U-238 in LEU fuels when compared with the HEU and further underscore the need to raise their thermal power by approximately 10% for efficient utilization of the MNSR in Neutron Activation Analysis.

Rapid quantitative elemental analysis using artificial neural network for case study of Isfahan Miniature Neutron Source Reactor

Rapid quantitative elemental analysis using artificial neural network for case study of Isfahan Miniature Neutron Source Reactor

SERPENT CODE POINT KINETIC PARAMETERS ANALYSIS: A NEXUS FOR NIGERIA RESEARCH REACTOR – 1 CORE CONVERSION USING ENRICHED URANIUM DIOXIDE FUEL

The International Atomic Energy Agency (IAEA) requires that all test and research reactors operating on Higher Enriched Uranium (HEU) should be converted to Low Enriched Uranium (LEU) for safety and security purposes. Nigeria having a Miniature Neutron Source Reactor (MNSR) has been long interested in fuel technological research not just to develop the area but also to meet with resolution on the nuclear treaty set out by the global nuclear regulatory body. In this study, reactor kinetic parameters such as effective delayed neutron fractions, prompt neutron lifetime as well as mean neutron generation time were analysed for Nigerian Research Reactor-1 (NIRR-1). Serpent Monte Carlo code 1.17 is used in the analysis. For delayed neutron parameters determination, we used fission probability iteration under one averaged generation time and neutron population rate.The calculated values for delayed neutron were recorded as analogue prompt and implicit prompt neutron lifetime, reproduction time and emission time are in the order of 3×10-7 (s), in agreement with the calculated data from the nuclear data libraries and some literature.The result for delay neutron fraction and other time-based parameters support the fuel core conversion for NIRR-1.The computational and pictorial results obtained from Serpent code simulation described well the transient behavior of the delayed neutron in this reactor.The analytical results also spelled out the relevance and compatibility of low enriched uranium dioxide fuel over higher enriched type.The result of this study conforms with other results obtained from similar reactors but with different Monte Carlo codes and with higher enriched uranium

Development of a polynomial model for power prediction of miniature neutron source reactors

Development of a polynomial model for power prediction of miniature neutron source reactors

Feasibility study of Isfahan MNSR as a calibration thermal neutron source

In the present work, the thermal beam-tube of Isfahan Miniature Neutron Source Reactor (MNSR) with known neutron dose-equivalent rate and flux is investigated as a calibration source of thermal neutrons. It is shown that 90% of the neutrons emerging from this tube are thermal. Measuring the response of TLD-600 and TLD-700 dosimeters exposed to these neutrons reveals that the both dosimeters respond linearly to neutrons in the range of 0.1 to 38 mSv. Moreover, after subtracting the contribution of mixed gamma rays using the calibration curves obtained in a standard 60Co gamma rays field, the response of TLD-600 dosimeter still behaves linearly with dose-equivalent. Finally, it is concluded that the dose-equivalent rate and flux of thermal neutrons in Isfahan MNSR are stable with time and this reactor can reliably be used as a calibration source for thermal neutrons.

Investigative study of the radiation damage on fuel clad of miniature neutron source reactor using computational tools

Core conversion requires some evaluation of the reactor safety. Changes to the reactivity worth, shutdown margin, power density and material properties are crucial to the proper functioning of the reactor. The focus of this article is to study the neutron flux distribution in the reactor core and radiation damage on candidate clads. The Ghana Research Reactor-1 (GHARR-1) operates at maximum power of 30 kW in order to attain a flux of 1.0× 1012 n·cm–2·s for the high enriched uranium core. Using the GHARR-1 core geometry, considering 348 fuel pins, the multiplication factor (Keff) is calculated at enrichments of 10%, 12.5%, 16%, 20%, 30% and 90.2%. The spectrum of neutron flux generated in the 26 group is also calculated at the specified enrichments. The ion/particle interactions with the targets (clad) were studied in the Stopping and Range of Ion in Matter code to establish the best clad material based on recorded defects and vacancies generated. From the calculations and simulations, the best choice from the candidate clads based on the assessment is SiC. The calculation of the fuel campaign length gives 7.5 years. The defects sustained by the prospective clad showed low susceptibility to swelling and other forms of deformation.