Syrian Miniature Neutron Source Reactor Research Articles

Analysis of Depletion and Inventory of the Fuel for the MNSR Research Reactor Using the Deterministic Code DRAGON5

The Syrian miniature neutron source reactor (MNSR), a 30-kW, 90.0% highly enriched uranium fueled (U-Al) MNSR-type reactor has gone critical. Under operating conditions of 2 h per day for 5 days a week at a peak thermal neutron flux of 1.0 × 1012 n/cm2·s, the estimated core life is 10 years. After the fuel is depleted, the full spent-fuel assembly will be replaced with new low-enriched uranium. This study presents the results of a multigroup fuel burnup and depletion analysis of the MNSR fuel lattice using the DRAGON5 transport lattice code. Furthermore, infinite multiplication factor k∞ and several two-group macroscopic parameters, including scattering cross section, fission cross section, total cross section, and diffusion coefficient, and the transport mean free path have been studied. In addition to this, fuel isotopic composition dependency on burnup was calculated as a part of this study. The results contained in this study can be used as a microscopic database for performing criticality safety analysis and shielding computations for the design of a spent-fuel storage cask for the MNSR reactor core.

Radiation damage estimation in the Al-alloy cladding of the MNSR reactor

The radiation damage rates in the Al-303-1-alloy cladding of the Syrian Miniature Neutron Source Reactor reactor has been numerically estimated with the The MCNP-4C and NJOY93 codes and the ENDF/B-VI library. The calculations showed that the Al-cladding alloy had received a maximum radiation damage rate equal to 7.01×10−9 (dpa/s). The total damage and helium production rates in the Al-cladding alloy were 0.13 (dpa) and 1.01×10−2 (appm, He), respectively. The contribution of the fast neutrons in the radiation damage was most effective.

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

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

Thermal hydraulic and safety analysis for core conversion (HEU–LEU) of Syrian Miniature Neutron Source Reactor

The paper presents the behavior and properties analysis of the low enriched uranium fuel compared with the original high enriched uranium fuel. The MNSR reactor core was modeled with both fuel materials and the reactor behavior was studied during the steady state and abnormal conditions. The MERSAT code was used in the analysis. The steady state thermal hydraulic analysis results were compared with that obtained from the experimental results hold during commissioning the Syrian MNSR. Comparison with experimental data shows that the steady-state behavior of the HEU core was accurately predicted by the MERSAT code calculations. The validated model was then used to analyze LEU cores with two proposed UO2 fuel pin designs. With each LEU core, the steady state and 3.77 mk rod withdrawal transient were run and the results were compared with the available published data in the literatures for the low enriched uranium fuel core. The results reveal that the low enriched uranium fuel showed a good behavior and the peak clad temperatures remain well below the clad melting temperature during reactivity insertion accident.

Calculation of the power distribution in the fuel rods of the low power research reactor using the MCNP4C code

Abstract The Monte Carlo method, using the MCNP4C code, was used in this paper to calculate the power distribution in 3-D geometry in the fuel rods of the Syrian Miniature Neutron Source Reactor (MNSR). To normalize the MCNP4C result to the steady state nominal thermal power, the appropriate scaling factor was defined to calculate the power distribution precisely. The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. The minimum power was found in the fuel ring number 9 and was 79.9 W. The total power in the total fuel rods was 30.9 kW. This result agrees very well with nominal power reported in the reactor safety analysis report which equals 30 kW. Finally, the peak power factors, which are defined as the ratios between the maximum to the average and the maximum to the minimum powers were calculated to be 1.18 and 1.31 respectively.

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.

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.

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

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

Design of a permanent epithermal neutron irradiation site in the Syrian MNSR

A permanent epithermal neutron irradiation site was designed in the Syrian Miniature Neutron Source Reactor (MNSR) by using cadmium as a thermal neutron shielding material. This site was designed by Cd-shielding the internal surface of the outer aluminum tube of the FOIS (First Outer Irradiation Site) in the MNSR. The MCNP-4C calculations showed that, to have a permanent epithermal neutron irradiation site for the ENAA using the cadmium, it is necessary to add the top beryllium shims of the reactor to compensate for the reactivity losses due to the neutrons absorption in the cylindrical cadmium shell. The activation detectors were used to measure the thermal and epithermal neutron fluxes in the FOIS. Distribution of the thermal neutron flux along the vertical direction of the outer irradiation capsule used in the FOIS has been found using MCNP-4C code, and experimentally by irradiating five copper wires. Good agreements were obtained between the calculated and the measured results.

Calculation and comparison of the neutron energy flux spectra in the Syrian MNSR irradiation sites using the MCNP-4C code

A 3-D neutronic model for the Syrian Miniature Neutron Source Reactor (MNSR) was developed earlier to analyze the reactor neutronic using the MCNP-4C code. This model was used in this paper to calculate the neutron energy flux spectra in the five inner and five outer irradiation sites of the MNSR reactor. The continuous energy neutron cross sections were evaluated from ENDF/B-VI library. The neutron fluxes were calculated using 69 energy groups. The neutron energy flux for each group was calculated dividing the neutron flux by the width of each energy group. The calculations showed that the distributions of the neutron energy flux spectra in the five inner irradiation sites were in good agreements. These results were noticed in the reactor outer irradiation sites as well. Using the neutron flux spectrum in the first inner irradiation site, the thermal (0.0–0.625 eV) and fast neutron fluxes (0.5–10 MeV) were calculated and found 1.035 × 10 12 and 3.022 × 10 11 cm −2 s −1 respectively. The measured fluxes in the first inner irradiation site were found previously to be (0.97 ± 0.06) × 10 12 and (2.89 ± 0.06) × 10 11 cm −2 s −1 respectively. Good agreements were noticed between the calculated and the measured results. These agreements verify the calculated neutron flux spectra in the reactor sites.

Passive nondestructive burnup monitoring of MNSR irradiated fuel by measuring photoneutrons produced within fission products

A passive nondestructive method for monitoring of Syrian miniature neutron source reactor (MNSR) fuel burnup is introduced. The inner irradiation site design inside the Be reflector was exploited to measure the generated photoneutrons induced by fission products hard gamma radiation in the subcritical state. The photoneutron flux was measured using gold foils as a function of cooling time and operation power. For cooling time ranges between 10 and 25 d, experiments show that 140Ba is the extremely dominating inducer of photoneutrons and the measured flux is proportional to the accumulated 140Ba. This result forms a new method base for MNSR fuel burnup monitoring. It might be used also as a safeguards technique to check the operator declared information.

Core conversion analyses of the Syrian MNSR reactor from HEU to LEU and MEU fuel with homogeneously mixed burnable poisons

A comprehensive analysis has been performed to investigate the conversion of the Syrian MNSR (miniature neutron source reactor) from current HEU fuel to selected alternatives LEU and MEU fuels. For this purposes the core design calculations related to design and engineering of LEU and MEU fuels have been carried out using the codes WIMSD/4 and BORGES-part of the MTR-PC and the code CITATION. Aiming at reducing the fuel enrichment by maintaining reactor power, thermal neutron flux and excess reactivity in the same range of the current MNSR design, two fuel alternatives of LEU (UO 2–Mg) and MEU (U 3Si x –Al) have been investigated. The results indicate that the first type (UO 2–Mg) realizes the criticality conditions with low enrichment of 20% using the similar overall design of the present HEU fuel pins, whereas the second type (U 3Si–Al) requires increasing the enrichment up to 33%. For the purpose of reactor core lifetime extension the possibility of mixing the burnable poisons Gd 157 and Cd 113 in the fresh core has been also explored. Thus, the calculation results indicate that the long-term control effect of Cd 113 on the excess reactivity is more homogeneous over the time due to the lower burn up rate of this burnable poison.

Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code

The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ( 197Au (n, γ) 198Au and 59Co (n, γ) 60Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

Estimation of the activation of local reactor shielding concretes

Many elements in reactor shielding concrete become radioactive due to interaction with neutrons during the operation of a nuclear reactor. These radioactive elements will build up over the time due to ( n, γ) reaction and contribute to the radioactive waste during decommissioning of the reactor, they will increase the dose from the inner part of the biological shielding (concrete) surrounding a nuclear reactor during maintenance works within reactor containment, and their effect should be taken in the calculations of the dose behind the concrete shield. Six mixtures of local ordinary and special shielding concretes have been investigated in this study. Their compositions were determined using X-ray diffraction and neutron activation analysis using the Syrian Miniature Neutron Source Reactor (MNSR) and consequently the dose rates. Based on the results of the analyses and comparing to the published limits in the references, it has been found that 60Co, 152Eu, 154Eu and 134Cs will be the most effective radioactive nuclides existing in the local concrete, but Dolomite aggregates contain the minimum limits of them and it has been found that Fe, Mg, and Ti the major elements existing in the local concrete.

Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions.

Assessment of cooling effects on extending the maximum operating time for the Syrian miniature neutron source reactor

Various schemes of cooling have been investigated for the purpose of assessing potential benefits on the operational characteristics of the Syrian MNSR reactor. A detailed thermal hydraulic model for the analysis of MNSR has been developed. The analysis shows that an auxiliary cooling system, installed in the pool which surrounds the lower section of the reactor vessel, will significantly offset the consumption of excess reactivity due to the negative reactivity temperature coefficient. Hence, the maximum operating time of the reactor is extended. Compared with experimental data, the suggested model proves to be valid for the analysis of MNSR behavior under both steady state and transient conditions.

Determination of the axial thermal neutron flux non-uniform factor in the MNSR inner irradiation capsule

A 3-D neutronic model, using the WIMSD4 and CITATION codes, for the Syrian Miniature Neutron Source Reactor (MNSR) is used to calculate the axial thermal neutron flux non-uniform factor in the inner irradiation capsule. The calculated result is 4%. A copper wire is used to measure the axial thermal neutron flux non-uniform factor in the inner irradiation capsule to be compared with the calculated result. The measured result is 5%. Good agreement between the measured and calculated results is obtained.

Estimating the four-factor product (ε p Pfnl Ptnl) for the accurate calculation of xenon and samarium reactivities in the Syrian Miniature Neutron Source Reactor

The modified 135Xe equilibrium reactivity in the Syrian Miniature Neutron Source Reactor (MNSR) was calculated first by using the WIMSD4 and CITATION codes to estimate the four-factor product (e p P fnl P tnl ). Then, precise calculations of 135Xe and 149Sm concentrations and reactivities were carried out and compared during the reactor operation time and after shutdown. It was found that the 135Xe and 149Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The 149Sm reactivities could be neglected compared to 135Xe reactivities during the reactor operating time and after shutdown.

Measurement of the fast neutron flux in the MNSR inner irradiation site

The WIMSD4 code was used to calculate the fast neutron flux spectrum and the fast neutron fission cross-sections for 238U, using six energy groups ranging from 0.5 to 10 MeV. These results, with the measured radioactivities of the 140Ba, 131I, 103Ru, 95Zr and 97Zr fission products emerging from the fission of the 238U foil covered with a cadmium filter, were used to measure the fast neutron flux in the Syrian Miniature Neutron Source Reactor inner irradiation site.

Prediction of the in-core power and the average core temperature using dose-rate measurements in the Syrian Miniature Neutron Source Reactor

The measurement of the dose-rate on the top-surface of the Miniature Neutron Source Reactor (MNSR) has been used to monitor the in-core reactor power and estimate the average core temperature. An experimental correlation for the prediction of both reactor power level and average core-coolant temperature as functions of dose-rate measurement have been obtained. An alternative method for both neutron flux monitoring and average core temperature using dose-rate measurement has been suggested. Comparing the results with the calibrated reactor power, reasonable agreement between predicted and real data measurement has been found. Despite of this agreement, the measurement of dose-rate on top of the MNSR surface as an additional mean for the prediction of in-core power or average core temperature should be considered carefully.