The program “MULTINAA” for various standardization methods in neutron activation analysis

Experimental determination and simulation of neutron and gamma flux parameters in horizontal channel for Prompt Gamma Neutron activation analysis implementation at Es-Salam research reactor

Since 2004, the k₀-standardization method has been applied and evaluated in three experimental channels at Es-Salam research reactor of the Algerian Centre de Recherche Nucléaire de Birine (CRNB). The NAA laboratory at CRNB is well involved in the implementation of the Prompt Gamma Neutron Activation Analysis to extend the field of analyses to some additional elements to the NAA method. In the frame work of the PGNAA research project, several works have been carried out in the present study including the experimental and simulation characterization of neutron and gamma flux parameters of the horizontal channel. The absolute and relative neutron fluxes of the studied channel have been determined by using a set of monitors including Al, Au, Cu, Fe and Zr. Furthermore, measurements of gamma and thermal neutron doses have been achieved using LiF: Mg, Ti (TLD 600 and TLD 700) detectors. The gamma dose rate at the experimental channel center position was found to be 0.6 Gy/min·MW. The experimental results obtained in this study have been used as input information for simulations and design of neutron facilities and radiation shielding, which are successfully established by SCALE code. The results were compared in terms of spectral detail and precision and a good agreement has been revealed. The value of the thermal neutron flux was found to be (6.29 ± 0.25) E+08 n/cm².s which allows the implementation of PGNAA at the Es-Salam research reactor.

Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia

A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k(0)-INAA and absolute method. The average values of φ(th),φ(epi), and φ(fast) by MCNP code were (2.19±0.03)×10(12) cm(-2)s(-1), (1.26±0.02)×10(11) cm(-2)s(-1) and (3.33±0.02)×10(10) cm(-2)s(-1), respectively. These average values were consistent with the experimental results obtained by k(0)-INAA. The findings show a good agreement between MCNP code results and experimental results.

Determination of fast neutron flux distribution in irradiation sites of the Malaysian Nuclear Agency research reactor

Determination of thermal to fast neutron flux ratio (f(fast)) and fast neutron flux (ϕ(fast)) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f(fast) and subsequently ϕ(fast) were determined using the absolute method. The f(fast) ranged from 48 to 155, and the ϕ(fast) was found in the range 1.03×10(10)-4.89×10(10) n cm(-2) s(-1). These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.

Removal of selenium from Se enriched natural soils by a consortium of Bacillus isolates

Four bacterial strains designated as SNTP-1, NS-2 to NS-4 were isolated from selenium contaminated soils of Nawanshahr-Hoshiarpur region of Punjab, India, by enrichment technique and a consortium was developed using these isolates. The isolates were observed to be belonging to Bacillus sp. In soil microcosm, complete removal was observed by the consortium in selenite augmented soils while the rate of removal with consortia in selenate treatment was 72% after 120 days. Population survival of isolates showed stability at lower treatments and decline at higher levels of Se enrichment. The consortium can, thus, be used for removal of Se contaminated sites.

Silicon doping system at the research reactor FRM II

Silicon doping has being carried out at FRM II since 2 years. During the commissioning of our new reactor, a simple test rig was used to determine the neutron flux profile at the irradiation position and optimise a nickel absorber liner, which is equipped at the irradiation position for vertical smoothing of the neutron flux profile. MCNP-code was used during the design of the liner. The final automatic doping system is designed to allow the irradiation of cylindrical silicon single crystals 500mm high and up to 200mm in diameter. Silicon ingots are additionally rotated continuously about their own cylinder axis during irradiation. The neutron flux density is measured online by using self-powered-neutron (SPN) detectors. The necessary doping homogeneity of +/-5% is achieved. The doping procedure and doping quality of ingots with high target resistivity are also discussed.

Production of 177Lu at the new research reactor FRM-II: Irradiation yield of 176Lu177Lu

Due to its physical and chemical characteristics, 177Lu is a very attractive radionuclide for use in nuclear medicine. This paper introduces a method for a precise calculation of the irradiation yield of 177Lu produced by neutron activation of 176Lu in a nuclear reactor. The calculation is based on the Westcott convention which requires the knowledge of the neutron flux parameters. In this work, the neutron flux parameters of the new research reactor FRM-II (Garching, Germany) were determined and the stability of thermal neutron flux and thermal neutron flux temperature was monitored. The comparison of theoretically calculated and experimentally determined yield for Lu176(n,gamma)Lu177 reaction is presented.

Validation of a neutron activation analysis method using k0-standardization

A neutron activation analysis method using k0-standardization has been adopted for multielement analysis using our research reactors. Gold has been used as the comparator. The input parameters, used for calculation of elemental concentration, such as subcadmium to epithermal neutron flux ratio (f), the epithermal neutron flux shape factor (alpha), and the absolute efficiency of the detector (epsilon) have been determined. The values of k0-factors and Q0 (the ratio of resonance integral to thermal neutron cross section) have been taken from the literature. The validity of applying this method in our laboratory is evaluated by analyzing the elemental concentrations with respect to the certified values in eight reference materials of different origin obtained from USGS, IAEA and NIST. The percent deviation of the measured elemental concentrations are found to be within +/- 15% to that of the certified values whereas the percent relative standard deviation ranged from 2% to 10% for most of the elements analyzed.