NUR Research Reactor Research Articles

Software-Based automation of burnup calculations for the NUR research reactor using SCALE/TRITON T6-DEPL sequence

This work represents a key milestone in the development of a Monte Carlo Burnup Calculation System (MCBCS) specially tailored for the NUR research reactor. Developed using Python and leveraging the 3D Monte Carlo TRITON depletion sequence (T6-DEPL) within the SCALE code, MCBCS accurately simulates the reactor’s operating history. The paper provides an overview of MCBCS, focusing on its components, verification, and validation.The verification and validation process cover both fresh and burnt core conditions. For the fresh core, comparisons of excess reactivity, control rods worth, and critical configurations against experimental data and MCNP5 calculations showed good agreement. Burnup calculations were validated against measured core excess reactivity, reactivity worths of fuel assemblies, and neutron flux distribution. The system slightly underpredicted core excess reactivity by −2.95%, and discrepancies in reactivity worths remained within the 7% uncertainty range. Neutron flux distribution showed good consistency with minor location-specific deviations.Overall, these findings confirm MCBCS as a reliable and accurate tool for burnup calculations of the NUR research reactor.

Performance evaluation and validation of OpenMC code for criticality analysis of an MTR-type research reactor

This paper assesses the performance of OpenMC, an open-source Monte Carlo code for particle transport, through criticality analysis of an MTR-type research reactor. The evaluation involves validation against experimental data from the NUR research reactor's first fresh fuel core and a code-to-code comparison with MCNP5, using a consistent methodology ensuring identical input conditions. Key parameters, including excess reactivity, control rod worths, shutdown margin, critical configurations, kinetics parameters, neutron flux, power distributions, and peaking factors, were examined. Results indicate good agreement between OpenMC simulations, experimental data, and MCNP5 results, with marginal differences well within acceptable ranges. Discrepancies with experimental data are below 5 %, considering measurement accuracy, while code-to-code comparisons reveal relative differences below 1 % in most cases. This study supports OpenMC's reliability for MTR reactor criticality analysis, affirming the validity of the developed NUR reactor model for future applications and highlighting the code's credibility and applicability in the nuclear reactor simulations.

Phthalonitrile resin reinforced silane surface modified boron carbide: Efficient neutrons screening and exceptional thermal properties

The design of lightweight neutron shields has been restricted for quite some time to the use of epoxy thermosets as the main building blocks. Meanwhile, the recent developments in the field of polymers suggest otherwise. Indeed, the phthalonitrile (PN) resins have taken the lead over traditional thermosets in many demanding applications. Therefore, in a vision to introduce newer matrices with better performances and to further expand the applications of the PN resins into the nuclear field, the neutron shielding efficiency along with the thermal resistance performances of the neat PN polymer and its subsequent silane surface-modified B4C-reinforced composites were investigated. The neutron shielding measurements were performed using an optimized experimental setup at the NUR research reactor in Algiers. The neat PN polymer displayed better thermal neutron screening performances than the epoxy and benzoxazine, with a macroscopic cross-section (Σ) of a 1.936 cm−1, equivalent to a mean free path (λ) of 0.358 cm. The effect of the particle amount was also studied to maximize the shielding ability of the developed materials. For instance, the PN composite containing 20 wt. % of B4C displayed an outstanding screening ratio of about 99.8% for a sample thickness of 13 mm. Finally, the remarkable findings were put into context by providing multifaceted comparisons with the available shielding materials.

Control Rod Worth Uncertainty Propagation from Nuclear Data for NUR Research Reactor

Control Rod Worth Uncertainty Propagation from Nuclear Data for NUR Research Reactor

Feasibility study of using the NUR research reactor for a BNCT installation and Monte Carlo optimization of a BSA

The aim of this work was devoted to study the feasibility of using the Algerian research reactor NUR for Boron Neutron Capture Therapy (BNCT) applications. To achieve this aim, an analysis of parameters and operational safety limits of the current core configuration X-1 with a removal of the thermal column was carried out. Furthermore, this study was followed by optimization calculations of the Beam Shaping Assembly (BSA) in the channel N∘5 using the Monte Carlo code MCNP5 simulations with a highlight of a new moderating material (Modified Fluental). Subsequently, a dosimetric study was performed using the MCNP5 code and the Snyder Head phantom by taking a Boron concentration of 65 ppm and 18 ppm in tumor and healthy tissue respectively. The in-phantom figures-of-merit have been calculated and provided promising results.

High Performance Dual Ballistic and Thermal Neutrons Shields From Kevlar Fibers Reinforced Epoxy/B4C Hybrid Composites

Targeting the development of advanced lightweight thermal and ballistic neutrons shields, a new hybrid composite was developed, for the first time, from Kevlar fibers, epoxy and boron carbide (B4C) particles. Kevlar fibers, as one of the strongest polymeric fillers, possess a high proportion of low Z atoms highly suitable for moderating neutrons. Meanwhile, these fibers can provide an additional efficient protection against high velocity projectiles. The B4C particles were added in various amounts, mainly for their excellent absorption of thermal neutrons. The nuclear shielding tests were performed at NUR research reactor using an optimized experimental setup. The obtained results confirmed the high shielding efficiency of all the developed materials. Meanwhile, the best performance was recorded at B4C amount of 20 wt.% with a macroscopic cross-section (Σ) of with a 3.638 cm−1 equivalent to a mean free path (λ) of 0.191 cm. The obtained results were compared with the most performant available shields and the data confirmed that superiority of the developed materials.

Outstanding thermal neutrons shields based on epoxy, UHMWPE fibers and boron carbide particles

Aiming the development of advanced and lightweight thermal neutrons shields, a new hybrid composite was developed from UHMWPE fibers, epoxy and boron carbide (B4C) particles. The UHMWPE fibres were chosen because of their high hydrogen contents and exceptional mechanical properties. The neutrons shielding tests were performed using an optimized experimental setup at NUR research reactor, Algiers. Overall, the developed materials displayed remarkable neutrons screening performances. Meanwhile, the best performance was recorded at B4C amount of 20 wt% with a macroscopic cross-section (Σ) of with a 0.313 cm−1 equivalent to a mean free path (λ) of 2.2 cm.

Benzoxazine resin as an interesting building block for advanced neutrons shields

The design of lightweight neutrons shields has been restricted for quite some time to the use of the epoxy thermosets as the main building blocks. Meanwhile, the recent developments in the field of polymers suggest otherwise. Indeed, benzoxazine resins have taken the lead over the traditional thermosets in many exigent applications. Therefore, in a vision to introduce newer matrices with better performances and to further expand the applications of the benzoxazine resins into the nuclear field, the neutron shielding efficiency along with the thermal and thermomechanical performances of the neat benzoxazine polymer and its subsequent B4C-reinforced composites were investigated. The neutron shielding measurements were performed using an optimized experimental setup at NUR research reactor, Algiers. The neat benzoxazine polymer displayed almost similar thermal neutrons screening performances than the epoxy with a macroscopic cross-section (Σ) of a 0.724 cm−1 equivalent to a mean free path (λ) of 0.957 cm. The effect of the particle amount was also studied to maximize the shielding ability of the developed materials. For instance, the benzoxazine composite containing 20 wt.% of B4C displayed the outstanding screening ratio of about 96% for a sample thickness of 13 mm. Finally, the remarkable findings were put into context by providing multifaceted comparisons with the available shielding materials.

Development of lightweight and highly efficient fast neutrons composites shields based on epoxy, UHMWPE fibres and boron carbide particles

In a vision to design wearable and lightweight protective thermal neutrons shields, a new hybrid composite was developed from Ultra High Molecular Weight Polyethylene (UHMWPE) fibres, epoxy and boron carbide (B4C) particles. The hybrid constituents were selected in a way to maximize both the shielding performances and the overall mechanical behaviour. Indeed, two first constituents of the hybrid attenuate the incident neutrons through scattering, while the B4C particles mainly interact with neutrons by absorption. More importantly, the UHMWPE fibres, as one of the strongest polymeric fibres, were also chosen to sustain the mechanical loads and provide the requested flexibility. The measurements were performed using an optimized experimental setup at NUR research reactor, Algiers. The effect of the amount of the B4C particles was also studied to maximize the shielding ability. The obtained results confirmed the high shielding efficiency of all the developed materials. Meanwhile, the best performance was recorded at B4C amount of 10 wt% with a macroscopic cross-section (Σ) of 0.188 cm−1 equivalent to a mean free path (λ) of 3.7 cm. In general, the developed materials performed better than the traditional neutrons shields and can be seen as promising materials for the intended use.

Effect of neutron irradiation on the structural, electrical and optical properties evolution of RPLD VO2 films

This study reports on the effect of neutron irradiation at different fluences on the properties of VO2 thin films. The irradiations were performed at NUR research reactor, Algiers at a temperature of about 40 °C, with fast neutron fluence (En > 1 MeV) up to 1.9 × 1018 n.cm−2. The induced defects have been investigated using structural, optical and electrical measurements. Both bulk sensitive characterization techniques, Raman and grazing incident angle X-ray diffraction (GIXRD) analysis, show that no structural transformation is induced by neutron irradiation, although strain induced defect production are generated throughout the films while surface sensitive techniques, X-ray photoelectron spectroscopy (XPS) and work function measurements, show that the charge carrier (electron) concentration at room temperature decreases after irradiation. Potentially due to fast neutron irradiation induced defects, mainly in the form of Frenkel pairs, swelling and color center formation occurs in VO2 thin films without amorphization. This is further corroborated by an increase of the room temperature resistivity through the irradiated films. Temperature-dependent electrical and optical transmission measurements confirm that the characteristic semiconductor-to-metal transition of the VO2 films is preserved upon irradiation. We therefore conclude that VO2 is an excellent candidate for thermal shielding and thermal management of small satellites.

Analysis of a loss-of-flow accident resulting from the primary pump shaft break transient of the NUR research reactor

ABSTRACT This paper presents a combined experimental and numerical simulation approach to investigating safety issues related to postulated loss-of-flow accident (LOFA) cases, which are more likely to occur in the NUR Research Reactor (Algeria). The transients investigated at nominal-power operating conditions are related to the loss of flow resulting from an instantaneous shaft break in the main cooling pump of the NUR reactor. The investigations are based on hydrodynamic and thermal hydraulic experiments to assess the reactor cooling system’s behavior. 3D Monte Carlo neutron transport calculations were performed with the (MCNP) code to determine the resulting neutronic properties of the core. In the accident analysis, a model of the primary cooling system was applied via the RELAP5 code. The experimental data and RELAP5 predictions showed good agreement. Additionally, the LOFA due to the transient scenario of the pump shaft break was compared with the LOFA due to normal loss of the coolant pump power transient. The results obtained from the transient (LOFA) studies revealed that in both cases, the lower limit of the minimum critical heat flux ratio and minimum onset of flow instability ratio for NUR is satisfied with a sufficient margin.

Neutron spectrum measurements at a radial beam port of the NUR research reactor using a Bonner spheres spectrometer

This paper describes the measurement campaign held around the neutron radiography (NR) facility of the Algerian 1MW NUR research reactor.The main objective of this work is to characterize accurately the neutron beam provided at one of the radial channels of the NUR research reactor taking benefit of the acquired CRNA Bonner spheres spectrometer (BSS). The specific objective was to improve the image quality of the NR facility. The spectrometric system in use is based on a central spherical 3He thermal neutron proportional counter combined with high density polyethylene spheres of different diameters ranging from 3 to 12in. This counting system has good gamma ray discrimination and is able to cover an energy range from thermal to 20MeV. The measurements were performed at the sample distance of 0.6m from the beam port and at a height of 1.2m from the facility floor. During the BSS measurements, the reactor was operating at low power (100W) to avoid large dead times, pulse pileup and high level radiation exposures, in particular, during spheres handling.Thereafter, the neutron spectrum at the sample position was unfolded by means of GRAVEL and MAXED computer codes. The thermal, epithermal and fast neutron fluxes, the total neutron flux, the mean energy and the Cadmium ratio (RCd) were provided. A sensitivity analysis was performed taking into account various defaults spectra and ultimately a different response functions in the unfolding procedure. Overall, from the obtained results it reveals, unexpectedly, that the measured neutron spectrum at the sample position of the neutron radiography of the NUR reactor is being harder with a predominance of fast neutrons (>100keV) by about 60%. Finally, those results were compared to previous and more recent measurements obtained by activation foils detectors. The agreement was fairly good highlighting thereby the consistency of our findings.

Radiation damage induced in Al2O3 single crystal sequentially irradiated with reactor neutrons and 90 MeV Xe ions

The present investigation reports the effect of 90MeV Xe ion irradiation on neutron irradiated Al2O3 single crystals. Three irradiation experiments were performed, with neutrons only, 90MeV Xe ions only and with neutrons followed by 90MeV Xe ions. Neutron and 90MeV Xe ion irradiations were performed at NUR research reactor, Algiers, Algeria and at GANIL accelerator, Caen, France respectively. After irradiation, the radiation damage was investigated by Raman spectroscopy, Fourier Transform Infrared Spectroscopy (FTIR), optical absorption measurements, and X-ray diffraction (XRD) techniques. Raman technique revealed that the concentration of the defects formed in Al2O3 samples subsequently irradiated with neutrons and 90MeV Xe ions is lower than that formed in Al2O3 samples which were irradiated only with neutrons. This reveals the occurrence of ionization-induced recovery of the neutron damage. Furthermore, as revealed by XRD analysis, a new peak is appeared at about 2θ=38.03° after irradiation at high fluence (>3×1013Xe/cm2). It can be assigned to the formation of new lattice plane.

ITHNA.SYS: An Integrated Thermal Hydraulic and Neutronic Analyzer SYStem for NUR research reactor

This paper introduces the Integrated Thermal Hydraulic and Neutronic Analyzer SYStem (ITHNA.SYS) that has been developed for the Algerian research reactor NUR. It is used both as an operating aid tool and as a core physics engineering analysis tool.The system embeds three modules of the MTR_PC software package developed by INVAP SE: the cell calculation code WIMSD, the core calculation code CITVAP and the program TERMIC for thermal hydraulic analysis of a material testing reactor (MTR) core in forced convection.ITHNA.SYS operates both in on-line and off-line modes. In the on-line mode, the system is linked, via the computer parallel port, to the data acquisition console of the reactor control room and allows a real time monitoring of major physical and safety parameters of the NUR core.PC-based ITHNA.SYS provides a viable and convenient way of using an accumulated and often complex reactor physics stock of knowledge and frees the user from the intricacy of adequate reactor core modeling. This guaranties an accurate, though rapid, determination of a variety of neutronic and thermal hydraulic parameters of importance for the operation and safety analysis of the NUR research reactor. Instead of the several hours usually required, the processing time for the determination of such parameters is now reduced to few seconds.Validation of the system was performed with respect to experimental measurements and to calculations using reference codes.ITHNA.SYS can be easily adapted to accommodate other kinds of MTR reactors.

NUR research reactor safety analysis study for long time natural convection (NC) operation mode

The current work represents a part of an overall effort being undertaken to improve NUR nuclear research reactor neutron utilization capabilities using a new core configuration. A RELAP5 model for the reactor NUR under natural convection (NC) operating mode has been developed. The model represents internal pool reactor components with the corresponding geometry, point neutron kinetics, and thermal hydraulics characteristics. An experimental device was designed and implemented in the reactor pool for monitoring the inlet and outlet core temperatures, and other pool temperature positions during NC-operating mode. In this paper, unprotected fast reactivity insertion and total flow blockage of the flapper valve transients have been investigated under NC operating conditions. The achieved steady-state results were found to be in good qualitative agreement with measurements. The results obtained from the transient (FRIA) study were compared to similar approach from recent literature. The second transient herein considered, is an attempt to predict the reactor core thermal-hydraulic behavior under a total flow blockage of the (NC) flapper valve. The latter could be considered as useful contribution for updating the safety analysis report.

BURNUR.SYS: A 2-D code system for NUR research reactor burn up analysis

Adequate knowledge of burn up levels of fuel elements within a research reactor is of great importance for its optimum operation. Such knowledge is required for the monitoring of reactivity parameters and flux and power distributions throughout the reactor core, the estimation of the radioactive source term needed in accidental situations analysis, the evaluation of the amount of fissile materials present at any moment within the fuel for safeguards purposes and the estimation of cooling and shielding requirements for interim storage or transport of spent fuel elements. This paper presents the approach of fuel burn up evaluation used at the NUR research reactor. The approach is essentially based upon the utilization of BURNUR.SYS code, an in-house developed software. BURNUR.SYS is an object oriented program under DELPHI 7 that integrates the cell calculation code WIMSD-4 and the core calculation code CITVAP. BURNUR.SYS calculates the evolution in time of pertinent quantities such as: the concentrations of U235 and others actinides, the concentrations of major poisons (Xe135 and Sm149), the distributions of power densities and burn up levels within fuel elements, the effective multiplication factor and core reactivity. The results are displayed in user friendly graphical and numerical formats.

Neutron flux optimization in irradiation channels at NUR research reactor

Optimization of neutron fluxes in experimental channels is of great concern in research reactor utilization. The general approach used at the NUR research reactor for neutron flux optimization in irradiation channels is presented. The approach is essentially based upon a judicious optimization of the core configuration combined with the improvement of reflector characteristics. The method allowed to increase the thermal neutron flux for radioisotope production purposes by more than 800%. Increases of up to 60% are also observed in levels of useful fluxes available for neutron diffraction experiments (small angle neutron scattering (SANS), neutron reflectometry, etc.). Such improvements in the neutronic characteristics of the NUR reactor opened new perspectives in terms of its utilization. More particularly, it is now possible to produce at industrial scales major radio-isotopes for medicine and industry and to perform, for the first time, material testing experiments. The cost of the irradiations in the optimized configuration is generally small when compared to those performed in the old configuration and an average reduction factor of about of 10 is expected in the case of production of Molybdenum-99 (isotope required for the manufacturing of Technetium-99 medical kits). In addition to these important results, safety analysis studies showed that the more symmetrical nature of the core geometry leads to a more adequately balanced reactivity control system and contributes quite efficiently to the operational safety of the NUR reactor. Results of comparisons between calculations and measurements for a series of parameters of importance in reactor operation and safety showed good agreement.

Safety analysis of neutron flux optimization in irradiation channels at the NUR research reactor

Prior to core reloading, planned power upgrading, or as a part of required analyses of past events, accurate safety evaluations should be carried out. Generally speaking, the content of a safety report has to be modified whenever a new type or design of fuel is to be used in a reactor core. As the existing plants have well established licensing procedures, including well founded analysis methods, the application of new analysis methods has to be thoroughly evaluated, with specific emphasis on their capability of producing results beneficial to reactor operation. The detailed study presented here was carried out so as to insure that the allowed operational safety limits of the NUR research reactor are not exceeded under any circumstances.

Analysis of resolution defects in the SANS spectrometer at the NUR reactor

The effects of the collimating system and of the dispersion in neutron wavelength on scattered intensities are reviewed in this paper for the SANS spectrometer at the NUR research reactor. It is found that the design parameters of the spectrometer will induce in general a non-negligible smearing effect in the measured (experimental) intensities. The introduction of adequate de-smearing schemes in the data treatment step is required in order to render the experimental data useful.

Improvement of performances of the small angle neutron scattering at NUR reactor

Small Angle Neutron Scattering (SANS) is a powerful technique for the investigation of inhomogeneities in matter. The technique provides valuable information over a wide variety of fields including chemical aggregation, defects in materials, surfactants, alloy segregation, polymers, biological membranes, viruses and macromolecules. In order to investigate the structures of micellar aggregations of asphaltene molecules as well as fluorinated surfactants which are of importance for the Algerian oil industry, a SANS spectrometer has been installed on beam port No. 4 of the NUR research reactor. Because SANS experiments are very time consuming and require low neutron background, an important effort was made to improve experimental conditions. This paper presents the general characteristics of SANS facility and addresses the various improvements that have been made.