Verification of Monte Carlo calculations of the neutron flux in typical irradiation channels of the TRIGA reactor, Ljubljana

An experimental verification of Monte Carlo neutron flux calculations in typical irradiation channels in the TRIGA Mark II reactor at the Joef Stefan Institute is presented. It was found that the flux, as well as its spectral characteristics, depends rather strongly on the position of the irradiation channel. Comparison of the results for parameter f (thermal-to-epithermal flux ratio), experimentally obtained by the Cd ratio multi-monitor method and Monte Carlo simulations shows relatively good agreement for most studied configurations except for the channel IC40 in the carousel facility, where differences are relatively large and not understood.     

New determination method of the neutron diffraction system resolution at the Algerian Es-Salam research reactor

The aim of this article is to determine the resolution of the neutron diffraction system (NDS) installed around the Algerian Es Salam research reactor. By using a new method based on neutron radiography technique we have measured the beam divergence at the exit of Soller Collimator (α₂) and hence determine effective collimations. The determination of the adequate resolution of our NDS is performed on experimental results obtained from neutron diffraction patterns for different collimation divergences (α₂) and (α₃) values obtained from several apertures sizes (S ₂) and (S ₃).     

Characterization of the neutron flux gradients in typical irradiation channels of a TRIGA Mark II reactor

The neutron distribution in a defined volume (gradient) for different matrices (air, water, cellulose, biological material and silicon dioxide) in two typical irradiation channels (pneumatic tube (PT) and IC40-channel in the carousel facility) in the TRIGA Mark II reactor at the Joef Stefan Institute (IJS) was studied. Our experiment was based on inserting Fe wires (flux monitors) into the chosen matrices. The wires were cut into small pieces after irradiation and the induced activities of ⁵⁹Fe measured. The results showed that for the studied geometry the average spatial thermal neutron flux inhomogeneities (for five studied matrices) are about 2.3% in the PT-channel and about 2.9% in the IC40-channel.     

The neutron flux of a252Cf neutron activation analysis device using the MCNP code

This report presents results from the application of the Monte Carlo N-Particle (MCNP) computer code to the²⁵²Cf neutron activation analysis (NAA) Device in the Technical Physics Institute of the Heilongjiang Science Academy of the People's Republic of China. The thermal and epithermal neutron flux at the sample positions and the neutron and photon fluxes on the surfaces of the device were calculated. A comparison between the calculated and experimental thermal and epithermal neutron fluxes at sample positions yield relative errors of less than 10% for the thermal neutron flux.     

Mapping of neutron flux gradient at NIST reactor to optimize neutron activation analysis of53Mn

Determination of⁵³Mn in meteorites by neutron activation analysis requires a thermal neutron flux high enough to ensure adequate production of⁵⁴Mn from⁵³Mn with a sufficiently low fast neutron component to minimize its production through fast neutron reactions. Thermal and fast neutron fluxes were mapped as a function of sample position within the NIST research reactor in order to determine the optimum position for irradiation of⁵³Mn.     

Measurement of neutron spectrum parameters for NAA irradiation holes in the Jordan research and training reactor

Three irradiation holes coupled to a pneumatic transfer system were installed for neutron activation analysis in the Jordan Research and Training Reactor (JRTR), which is the first research reactor in Jordan. To perform instrumental neutron activation analysis, neutron spectrum parameters, such as thermal neutron flux, α and f for the irradiation holes, should be measured. The Cd-ratio method was applied for the determination of the aforementioned parameters. For this purpose, 0.1% Au–Al wires and Zr foils were irradiated with and without Cd-cover, and the Cd ratios were determined for Au-198, Zr-95, and Zr-97/Nb-97m nuclides. Then, the parameters were calculated and determined at three irradiation holes.

     

Variation of neutron flux and related parameters in an irradiation container,in use with k0-based neutron activation analysis

Profiles of thermal flux, epithermal flux, f and have been determined for a multisample irradiation container in use with k₀-based NAA. It could be concluded that for accurate NAA with the IRI-reactor, samples should be surrounded by at least two thermal flux monitors. Since f and values were found to vary considerably between irradiations as compared to the variations within one container, the possibility of determining f and just once for the irradiation facility used, and monitoring only the thermal flux in future irradiations, was ruled out. The use of average values for f and within a container was shown to yield unacceptable inaccuracies mainly for Zr and U determinations.     

Determination of neutron temperature in irradiation channels of reactor

The neutron temperature is a characteristic parameter in irradiation channels of reactor. For nuclides which have resonances in the thermal neutron energy range, their Westcott g-factors are different from unity. The values of g-factors and cross-sections of (n, γ) reaction of these nuclides are temperature dependence. The standard energy for tabulation of thermal neutron cross-section (σ₀) is that of room temperature (293.59 K or 20.43 °C), corresponding to a neutron energy 0.0253 eV or to a neutron velocity of 2200 m/s. However, in the irradiation channels of reactor, the temperature is not exact at 20.43 °C. Thus, the temperature at the irradiation position must be known to convert σ₀ to σ(T). A method for determination of the neutron temperature in irradiation channels of Dalat reactor is presented by fitting the thermal neutron spectrum obtained from the calculation using MCNP code.     

Measurement of the neutron flux distribution in a research reactor by neutron activation analysis

Journal of Radioanalytical and Nuclear Chemistry - The neutron flux distribution in the core region of the JRTR was measured through the neutron activation analysis method. Sample mount and a tool...     

The epithermal neutron flux distribution in a nuclear reactor and its effect on epithermal neutron activation analysis

The shape of the epithermal neutron energy distribution has been determined in two irradiation positions of the University of London CONSORT II reactor. The method applied involves cadmium ratio measurements using a series of resonance detectors. Principles of the method and some considerations relative to epithermal neutron activation analysis in connection with the deviation of the epithermal neutron flux distribution from the 1/E law are given.     

Determination of characteristic neutron flux parameters of nuclear reactors for thermal neutron activation analysis

The characteristic neutron spectrum parameters for thermal neutron activation analysis have been determined for the most important irradiation positions of the reactors BR1 and BR2 at Mol (Belgium), HFR at Petten (Netherlands) and FRJ2 at Jülich (Federal Republic of Germany). The method of determination is described.     

Calibration of the irradiation channel DBVK at BER-II reactor and feasibility of applying the k0-method at this device

This work is about k ₀-INAA using unstable neutron flux for sample irradiation. Due to slow transport, each irradiation in the channel DBVK consists of three phases: stable irradiation at the final position, and two additional irradiations during travelling by exposure to an increasing neutron flux in the delivery course and to a decreasing neutron flux in the fetch course. In this work, the neutron flux distribution along this channel was calibrated and the neutron flux variation with irradiation time was calculated, making it possible to evaluate activity growth during a complete irradiation period. The feasibility of the k ₀-method was checked by analyses of four SRM-materials and three multi-element standards at three DBVK-positions. An accuracy of better than ±10% was found for nearly all determined elements in each determination.     

Reevaluation of neutron flux characterization parameters for NIST RT-2 facility

It has been difficult to characterize the thermal to epithermal neutron flux ratio (f) and the measure of the nonideal epithermal neutron flux distribution (α) for the RT-2 pneumatic rabbit facility at the NIST National Bureau of Standards Reactor (NBSR). In a previous paper, only cadmium-covered irradiations yielded physically reasonable parameters. New measurements were performed using chromium, manganese, cobalt, zinc, zirconium, molybdenum, antimony, gadolinium, lutetium, and gold. The neutron temperature (T _n ) in RT-2 measured using bare lutetium and gold foils gave unphysical values. The bare foil methods for measuring f and α gave inconsistent results. The underlying reasons are demonstrated via MCNP simulation results for cumulative reaction rates of selected isotopes. To determine expected intervals for f, α and T, parametric methods were explored. Measured reaction rate probability per target atom (R _p ) values for the listed elements were fitted to a modified Westcott curve using an iterative least-squares method to verify consistency of measurements and nuclear data. An advanced parametric approach using a detailed MCNP model of the NBSR was used to calculate neutron flux characterization parameters.     

Reactor power dependency of the neutron flux parameters fork 0-standardization

Using the facilities of the Triga Mark III reactor at the NNRI, Mexico and the HAV-1 multipurpose monitor, the reactor power dependency for thek ₀-standardization essential neutron flux parameters as: epithermal shape factor (), thermal to epithermal ratio (f) and neutron temperature (T _n ) were experimentally obtained. Evaluation of the obtained dependencies shows that it is unnecessary to analyze the possible introduction of correction factors in thek ₀-INAA experimental results. A single experimental procedure to determine throughf is suggested.     

Estimation of background interference in prompt-gamma neutron activation using MCNP

Prompt-gamma neutron activation (PGNA) is used to measure total-bodynitrogen and hydrogen in humans. Background interference in the gamma spectraarises from both subject and shielding. A Monte Carlo simulation program (MCNP4B2)was used to examine the neutron and gamma signals in the PGNA system ( ²⁴¹AmBe source). N and H peak regions were assessed in the presenceand absence of calibration phantoms. The simulations suggested extracorporealH peak contributions of up to 30%, depending on subject body habitus. MostN background could be attributed to detector pileup events. The MCNP resultsallowed us to improve shielding design and develop background correction algorithmsto improve measurement precision.     

The slowing-down thickness dependency of some neutron flux parameters fork 0-standardization

The experimental dependence of the α,f andT _n parameters, in function of the water thickness, for different irradiation channels of Triga Mark III reactor, were analyzed. An exponential law for the α(r) dependence was obtained in the neighborhood of the active zone of the reactor numerically modelated using theS _n method for 69 neutrons groups, and this dependence is slower in light water reactors than in graphite reactors.     

Simulation and analysis of neutron energy spectra from irradiation channels of the IBR-2 reactor

With the purpose of perfecting neutron activation analysis calculation methods, neutron energy spectra for three irradiation channels were modeled and analyzed. On the basis of the results obtained, the renormalization factors were obtained for resonance integrals of the real resonance neutron spectrum from one of the irradiation channels.     

Determination of neutron flux parameters in PUSPATI TRIGA Mark II Research Reactor, Malaysia

In standardization NAA, it is necessary to characterize the neutron spectrum parameters such as epithermal neutron flux shape factor (α), thermal to epithermal neutron flux ratio (f), thermal neutron flux (φ _th) and epithermal neutron flux (φ _epi) in the irradiation facility to determine the concentration of an element in the sample using absolute and k ₀ standardization methods. The α and f were determined using Cd-ratio multi monitor method using experimental data obtained in PUSPATI TRIGA Mark II research reactor at four irradiation positions (10, 20, 30 and 40) of the rotary rack. The calculated values of α and f ranged from 0.006 to 0.0281 and 18.56 to 19.12 respectively. The average values of φ _th and φ _epi were found as 2.33 × 10¹² and 1.23 × 10¹¹ n cm^?2 s^?1 respectively. Moreover, a comparison of the neutron flux parameters in the present study shows an acceptable level of consistency with those of previous studies.     

Design and testing of a neutron porosity probe using MCNP code

A neutron porosity probe has been designed. The MCNP Monte Carlo code has been utilized to determine the various parameters of the probe. The probe response was simulated and compared with the experimental results with good agreement. Preliminary fieldwork was performed with acceptable data. This revised version was published online in July 2006 with corrections to the Cover Date.