The application of a252Cf neutron multiplier for routine instrumental neutron activation analysis

The successful application of instrumental neutron activation analysis for routine determinations depends on the ability to produce accurate and precise analytical results in a relatively short time. An important factor in obtaining the desired speed has been the availability of a low-cost, moderate-flux neutron source for on-site use. The²⁵²Cf neutron multiplier (CFX), designed and constructed by Intelcom Rad Tech Corporation of San Diego, California, is a subcritical assembly capable of continuous, stable operation and has provided us with the ability to determine more than 35 elements as major and minor components. The CFX produces a thermal neutron flux of ∼2×10³ n/cm²-sec by a 100-fold multiplication of the neutrons emitted from a 1 mg²⁵²Cf source. Of particular importance in its application at Kodak has been the determination of the halogens Cl, Br, and I, both singly and simultaneously, in various matrices including photographic materials.     

Neutron activation analysis with PE moderated252Cf neutron source

The 1.369 MeV -rays emitted from²⁴Na²⁴Mg+^–+ after²³Na(n,)²⁴Na reaction were counted by high purity germanium (HPGe) detector and the half-life of²⁴Na was derived. This process was simulated by Monte Carlo Neutron and Photon Transport Code (MCNP-4A).²⁵²Cf neutron source was moderated by a polyethylene (PE) cylinder to increase the cross section of neutron absorption reaction and to decrease the biological hazard. NaCl powder of 20 cm³ and 40 cm³ volume in cylindrical polypropylene capsules were irradiated by the neutrons passed through the moderator. MCNP-4A was used to determine the optimum size of PE moderator, to assume the realistic geometry of the HPGe detector, and to assume the absolute efficiencies of the detector. The count rates for 1.369 MeV -rays in the HPGe detector were calculated by MCNP-4A for 20 cm³ and 40 cm³ NaCl. The accumulated counts calculated tumed out to be higher than those actually measured by 31% with a relative error of 3%. The half-life of²⁴Na measured within 4% and 1% for 20 cm³ and 40cm³ NaCl agrees with that of the reference. So, we can say that the result of MCNP-4A has about 30% of accuracy and 3% of precision in simulating the neutron activation analysis.     

Activation analysis of halogens in photographic emulsions using a252Cf neutron multiplier

A nondestructive neutron activation technique for the simultaneous direct determination of chlorine, bromine, and iodine in silver halide mixtures is described using a²⁵²Cf neutron multiplier (CFX). About 5–50 mg of sample are used. The analysis is semiroutine, without the need for a monitor included with each sample since the flux stability and reproducibility are within ±1%, decaying only with the 2.65 year half-life of²⁵²Cf. The precision and accuracy are counting-statistics controlled and are generally ±1% RSD or better. The method offers an attractive alternative to existing chemical and instrumental methods for these determinations in silver halide mixtures because it has the potential for providing reasonably rapid analyses with good precision and accuracy.     

Fabrication of 50-mg252Cf neutron sources for the FDA activation analysis facility

The Transuranium Processing Plant (TPP) at the Oak Ridge National Laboratory (ORNL) has been requested by the Food and Drug Administration (FDA) to furnish 200 mg of²⁵²Cf for use in their new activation analysis facility. This paper discusses the procedure to be employed in fabricating the californium into four neutron sources, each containing a nominal 50-mg of²⁵²Cf. The completed neutron sources will be assayed using a precision fast-neutron counter, decontaminated and loaded into a concreete-shielded shipping container weighing 10.7 Mg for shipment to the FDA facility located at Howard University in Washington, D. C.Research sponsored by the Office of Basic Energy Sciences, U. S. Department of Energy, under contract DE-AC05-840R21400 with Martin Marietta Energy Systems, Inc.     

Instrumental neutron activation analysis of Al,V and Ti employing252Cf as a thermal neutron source

A rapid method for the determination of Al, V and Ti has been developed and is used for the analysis of these elements in different ores and alloys. An isotopic neutron source²⁵²Cf having a thermal neutron flux of the order of 8.5×10⁷ n·cm^–2 sec^–1 has been used for thermal neutron bombardment. Activity measurements were performed on a HPGe detector coupled to a PC based MCA unit. Depending on the half-life of the (n, ) product, different irradiation and cooling times were employed and thus the elements of interest were analyzed sequentially.     

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.     

A neutron induced prompt gamma-ray spectroscopy system using a 252Cf neutron source for quantitative analysis of aqueous samples

A neutron induced prompt γ -ray spectrometry (NIPS) facility has been developed at the Nuclear Chemistry Research Division, of the Korea Atomic Energy Research Institute (KAERI) with the aim of analyzing the major components of various elements in aqueous samples. The facility is equipped with a ²⁵²Cf neutron source and a γ-γ coincidence setup with two n-type coaxial HPGe detectors based on NIM spectrometric modules in association with data acquisition and spectral analysis systems. The development of the system, its set-up and the calibration of detection efficiency up to 8 MeV using a set of radionuclides and the (n,γ) reactions of chlorine are described in the paper. This revised version was published online in July 2006 with corrections to the Cover Date.     

Determination of some macro and micro nutrients in soils of Maharashtra State by instrumental neutron activation analysis using 252Cf as a thermal neutron source

Some macro and micronutrients in soils from some districts of Maharashtra State have been determined by instrumental neutron activation analysis. Neutrons from ²⁵²Cf have been used for irradiation. The activity induced was measured by a HPGe detector coupled to a MCA unit. The statistical evaluation of the method has been explained. This revised version was published online in August 2006 with corrections to the Cover Date.     

In vivo analysis for nitrogen using a252Cf source

A set up forin vivo determination of nitrogen has been built. Phantoms containing different amounts of nitrogen have been measured as well as a volunteer in a pilot study. A total body protein content of 18.8 kg was calculated, to be compared with 17.0 kg estimated from potassium measurements.     

In situ activation analysis of marine sediments with252Cf

A seabed nuclear probe has been developed which permits the measurement of up to 33 elements at the parts per thousand to parts per million levels in continental shelf areas. The probe consists of a neutron irradiation source,²⁵²Cf, which neutron activates the elements in the minerals of interest. These resulting radioactive elements emit characteristic gamma radiation which is analyzed in situ in 2- to 10 min counting intervals with a Ge(Li) detector system. Details of the irradiation source, the Ge(Li) detector, and spectra taken in situ in a marine environment are discussed with respect to the detection of the minerals at concentrations of economic interest.     

Determination of manganese in ores by activation analysis using a californium-252 neutron source

A²⁵²Cf neutron source has been used to analyse manganese in ores such as pyrolusite, rodonite (manganese silicate) and blends used in dry-batteries. Samples with about 150 mg and standards of manganese dioxide were irradiated for about 20 min and counted using a well-type NaI(Tl) scintillation counter and scaler, with or without pulse-height discriminator between the detector and the scaler. The interferences of nuclear reactions⁵⁶Fe(n,p)⁵⁶Mn and⁵⁹Co(n,α)⁵⁶Mn were studied, as well as problems in connection with neutron shadowing during irradiation, gamma-rays attenuation during counting and influence of granulometry of samples. Some of the samples were also analysed by wet-chemical method (sodium bismuthate) in order to compare results.     

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.     

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...     

Monitoring of neutron flux changes in short-lived neutron activation analysis

It is well known variations in neutron fluxes can adversely affect the final result in neutron activation analysis. The monitoring of neutron flux changes are usually described for medium and long-lived NAA using foils of cobalt, gold, zirconium, etc. However, for short-lived neutron activation analysis there appears to be no systematic study of the variations of the neutron flux. With our new automatic pneumatic system, where irradiation timing, decay and counting and position are very reproducible, we have performed a series of experiments using thermal and epithermal neutrons using aluminum wire as a monitor to monitor the neutron fluxes. Our experiments confirm that neutron flux fluctuations in the worst case can be up to ±12 % with a SD of 2–3 %. This effect can be seen regardless of the irradiation time and must be taken into consideration to achieve the best result.     

Study of a neutron radiography system using Cf neutron source

This paper presents a study about moderation and collimation of a neutron radiography system using ²⁵²Cf. A Monte Carlo Code, MCNP4B, has been used to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet next to the image plane. Among the various moderator materials investigated, high density polyethylene proved to be the most efficient, with a thermalization factor of 56 cm². Using a collimator design assembly it was possible to obtain a normalized thermal neutron flux, at the image plane of 6×10⁻⁶ n cm⁻² s⁻¹ at an effective collimator ratio of 7.5, or 3.2×10⁻⁷ n cm⁻² s⁻¹ at an effective collimator ratio of 50.     

Use of252Cf for elemental analysis

The neutron sources based on the²⁵²Cf isotope and having relatively long half-lives and the highest specific neutron yields compared with other ampouled sources are proved to be the most convenient for use in industrial analytical laboratories situated far from the basic nuclear centers. Research making use of californium sources is currently carried out in a number of nuclear centers of the USSR, including the Institute of Nuclear Physics of the Uzbek Academy of Sciences, where a great amount of work have been done on verification of practical use of Cf sources of various activity in different branches of industry, first of all, in geology and mining industry.     

Accounting for flux decrease in neutron activation analysis

The problem of neutron flux decrease during the activation of samples is discussed. Data by different authors on the weight of spherical samples corresponding to a flux decrease of 10% are reported. The self-shielding effect due to the presence of constituents with large absorption cross-sections is discussed.     

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.