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The paper describes verification of the MCNP model of manganese bath used at the Czech Technical University for education purposes. The bath consists of a spherical tank made of plexiglass filled with an aqueous solution of MnSO 4 with the total volume of 163.3 litres. Neutron sources are introduced into the centre of the bath manually using a dry plexiglass channel. The MCNP model contains all important geometric and material details of the bath including optional neutron source. Its verification was performed based on experimentally determined activity; as well, neutron emission rates for used radionuclide sources were compared. The experiments were performed with two various radionuclide neutron sources (Am-Be and 252Cf). Comparison of calculation results with ENDF/B-VII.1 and JEFF-3.2 nuclear data libraries was performed.

Time required for activation to saturation and background measurement is considered a limitation of strength determination of radionuclide neutron sources using manganese bath system (MBS). The objective of this research was to evaluate the other solutes based on water bath for presentation of the suitable replacement with MBS. With the aid Monte Carlo simulation, for three neutron sources, having different neutron spectra, immersed in six aqueous solutions, i.e., Na2SO4, VOSO4, MnSO4, Rh2(SO4)3, In2(SO4)3, I2O5, the correction factors in all nuclei of solutions for neutron losses with different process were obtained. The calculations results indicate that the Rh2(SO4)3 and VOSO4 are best options for replacing with MnSO4.

In this work, the Monte Carlo method has been used to simulate the manganese sulfate bath calibration system, in order to evaluate the different interaction of neutron with all nuclei of the components of the solution, i.e., hydrogen, manganese, sulfur, and oxygen nuclei for 241Am–Be and 252Cf neutron sources. Also, the leakage probability of neutron has been calculated for various radiuses of spherical bath containing a solution with weight concentration of 93.68 g/(kg of solution). The simulations performed by MCNPX, included a detailed description of the geometry and material of the system (tank, holder and source). The calculations were evaluated using neutron cross sections from several libraries. The results of calculation show that the proper radiuses of spherical manganese bath with this concentration, for 252Cf and 241Am–Be sources, are about 35 and 60–65 cm, respectively.

A detailed CAD model of the new manganese bath facility of the UK National Physical Laboratory has been converted into MCNPX format using the conversion program MCAM. The model has been used to evaluate a number of important parameters of the new facility such as the fraction of neutrons that escape from the manganese sulphate solution but subsequently return and are captured by it, and the degree of irradiation of the solution when a source is being transferred to and from the bath. It has also been used to recalculate the correction factors for the fraction of neutrons captured by nuclei other than manganese, the fraction captured by the source and the source mounting assembly and the fraction escaping from the bath itself, and compare these with values from an earlier simplified model.

An accurate measurement of the 4.438 MeV gamma-ray to total neutron ratio, namely R=Sgamma/Sn, for a Chinese-made Am-Be neutron source is described. The neutron strength of the source relative to a previously standardized source was determined by the manganese bath technique. The gamma-rays spectra of the source were measured using a Ø 75 x 75 mm NaI(Tl) detector. The background induced by neutrons and the absolute full energy peak efficiency of the detector were calculated using the MCNP code. The experimental ratio so obtained agrees well with the calculated value. A synthetic evaluated and recommended value of R=0.575+/-4.8% was given. We conclude that the experimental R-value appears an important characteristic for an Am-Be source.

This work deals with the absolute measurement of the neutron emission rate from a 241Am-Be source by means of the manganese sulphate bath technique, which is the principal method for the absolute determination of the neutron emission rate from radionuclide neutron sources. The facility consists of a spherical container filled with an aqueous solution of manganese sulphate with a 241Am-Be neutron source placed at the center. As well known, neutrons from the source, after having been thermalized by the aqueous solution, undergo neutron capture by hydrogen, manganese, sulphur, and oxygen nuclei, thus inducing a certain activity to the solution. Subsequent gamma spectrometry measurements of 56Mn activity generated by 55Mn neutron activation allows to determine the neutron emission rate of the source, The experimental activity has involved a variety of measurement techniques and calculation procedures, ranging from neutron reactor activation to liquid scintillation counting and Monte Carlo calculations. Neutron activations of 55Mn samples has been carried out with the TRIGA reactor of the ENEA-Casaccia Research Centre, and 56Mn activated samples were subsequently characterized by liquid scintillation counting, in order to obtain reference standards for the calibration of the NaI(Tl) scintillation detectors utilized to record gamma-ray emission from 56Mn. Monte Carlo calculations, carried out by the MCNPX code, were required to calculate neutron transport within the sulphate manganese bath, in particular to determine 55Mn neutron capture probability, and (n, α) and (n, p) concurrent reactions, as well as the neutron leakage. Such a procedure has allowed to maintaining the neutron emission rate uncertainty well below 1 %. All the measurements have been carried out at the ENEA-Casaccia Research Centre by the Italian National Institute of Ionizing Radiation Metrology. © 2014 Akadémiai Kiadó.

Section III (neutron measurements) of the Comite Consultatif des Rayonnements Ionisants, CCRI, conducted a key comparison of primary measurements of the neutron emission rate of an 241 Am-Be(α,n) radionuclide source. A single 241 Am-Be(α,n) source was circulated to all the participants between 1999 and 2005. Eight laboratories participated – the CIAE (China), CMI (Czech Republic), KRISS (Republic of Korea), LNMRI (Brazil), LNE-LNHB (France), NIST (USA), NPL (UK), and the VNIIM (Russian Federation) – with the NPL making their measurements at the start and repeating them near the end of the exercise to verify the stability of the source. Each laboratory reported the emission rate into 4π sr together with a detailed uncertainty budget. All participants used the manganese bath technique, with the VNIIM also making measurements using an associated particle technique. The CMI, KRISS, VNIIM, and later the NPL, also measured the anisotropy of the source although this was not a formal part of the comparison. The first draft report was released in May 2006 and having been discussed and modified by the participants and subsequently reviewed by the CCRI(III), the present paper is now the final report of the comparison.

Two methods of analysing the results from the continuous-flow manganese bath are described, one enabling it to be used for time-varying neutron sources and the other making it possible to predict both the steady background and that due to activity present in the system initially. The results of tests on these methods are described, and the implications with regard to greater flexibility of operation are discussed.

Exact and approximate solutions of the equation governing the activity growth in continously cycled assemblies are presented. These solutions are applied to manganese and vanadium systems of practical interest and the implications for neutron yield measurements are discussed.

An accurate measurement of the gamma to neutron ratio for an Am-α-Be neutron source is described. The total neutron output was measured using an improved manganese bath technique and the 4.43 MeV gamma ray associated with the neutron producing reaction was measured absolutely with a NaI(Tl) detector. The experimental ratio so obtained agrees well with the theoretically computed value. It is recommended to use the experimental value of this ratio, namely 0.56±6%, as it is traceable to the primary standards.