Experimental studies of fission product release from model fuel elements at the physical start-up of the IVG.1M research reactor.

During the physical start-up of the IVG.1M research reactor, which involved the conversion to low-enriched uranium fuel, various studies were conducted using physical mock-ups of the fuel assemblies. The mock-up consisted of 468 model low-enriched uranium fuel elements that were cut into three sections without end coatings at the cut locations. This configuration allowed for the simulation of fuel element decompression and the experimental determination of fission product release into the coolant. The coolant samples obtained from the mock-up were analyzed using a gamma-spectrometer equipped with a high-purity germanium detector. As a result of the studies, the activities of fission products in the coolant and the averaged relative fission product release were determined. The relative release is defined as the ratio of the release rate of reference fission products to its birth rate. For the first time, fission product release was experimentally obtained in a situation simulating the decompression of IVG.1M fuel cladding.

Neutron spectroscopy with TENIS using an artificial neural network.

In this research, a ThErmal Neutron Imaging System (TENIS) consisting of two perpendicular sets of plastic scintillator arrays for boron neutron capture therapy (BNCT) application has been investigated in a completely different approach for neutron energy spectrum unfolding. TENIS provides a thermal neutron map based on the detection of 2.22 MeV gamma-rays resulting from 1H(nth, γ)2D reactions, but in the present study, the 70-pixel thermal neutron images have been used as input data for unfolding the energy spectrum of incident neutrons. Having generated the thermal neutron images for 109 incident mono-energetic neutrons, a 70 × 109 response matrix has been generated using the MCNPX2.6 code for feeding into the artificial neural network tools of MATLAB. The errors of the final results for mono-energetic neutron sources are less than 10% and the root mean square error (RMSE) for the unfolded neutron spectrum of 252Cf is about 0.01. The agreement of the unfolding results for mono-energetic and 252Cf neutron sources confirms the performance of the TENIS system as a neutron spectrometer.

Experimental studies of power distribution in LEU-fuel of the IVG.1M reactor.

For the past ten years, the IVG.1M research reactor has been converted from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Currently, the final stages of the IVG.M reactor conversion, including physical and energy start-ups, are underway. As part of the physical start-up, one of the objectives was to conduct physical studies on the power distribution across the radius and height of the fuel assemblies in water-cooled technological channels. Studies were carried out using the activation gamma-spectrometric method based on correspondence between energy release and measured gamma-radiation activity of fission or activation products. The ENREDI program was used to determine the detailed relative power distribution over radial cross section of fuel assembly. During these experimental studies, relative power distribution along the height and radial cross section of the IVG.1M fuel assembly, as well as power peaking factors were obtained. These studies will make it possible to evaluate the changes in the IVG.1M reactor fuel power profile after the fuel enrichment reduction, and to verify the accuracy of neutronic simulations.

Methods to study power density distribution in the IVG.1M research reactor after conversion.

In the Republic of Kazakhstan, within the framework of international initiatives on the non-proliferation of nuclear weapons, the IVG.1M research reactor is being converted to low enriched fuel. Due to the change in the fuel composition, it becomes necessary to assess the power density distribution in the fuel assemblies and in the core altogether. Close attention to this parameter is explained by the fact that the non-uniformity of the radial and axial power density fields determines the operating parameters of the reactor, its power, the core resource and the efficiency of fuel utilization. The purpose of this work is approbation of the activation gamma-spectrometry method for determining the power density distribution in the fuel of the IVG.1M research reactor after conversion. Additionally, for a preliminary assessment of the experimental outcomes, the paper presents the MCNP6 calculation results for the power distribution in fuel assemblies and the peaking factors.

Conceptual design for a new heterogeneous 241Am-9Be neutron source assembly using SOURCES4C-MCNPX hybrid simulations.

In this study, an approach to simulate a novel variable-yield heterogeneous 241Am-9Be was proposed with a hybrid use of SOURCES4C and MCNPX codes, and its energy spectrum and neutron emission yield were simulated. In these simulations, the energy spectra of the alpha particles emitted from the americium source and the neutrons produced within the beryllium and oxygen contents as a result of 9Be(α,n) and 17,18O(α,n) reactions were calculated with SOURCES4C whilst the neutron transport from neutron production points to the space outside the source assembly were performed with the MCNPX code. The neutron energy spectrum and neutron emission yield for two different configurations of single-rod and multi-rod sources (i.e., americium or americium oxide rods in beryllium medium) were compared to a source of homogeneous americium (or its oxides) and beryllium mixture. The proposed heterogeneous geometry was aimed to provide a neutron source with a variable neutron yield, easy-to-shut down and easy-to-waste process features. The results confirmed that the homogeneous source represented the largest neutron yield compared to single- and multi-rod geometries. However, the neutron yield in heterogenous geometry could be altered by varying the number of americium (or americium oxide) rods to reach the desired neutron yield.

A fuel for generation IV nuclear energy system: Isotopic composition and radiation characteristics.

This paper reports on an important issue of designing appropriate nuclear fuel of a high-temperature gas-cooled nuclear reactor operating in a thorium-plutonium nuclear fuel cycle. The neutronic calculations for a fuel of specific isotopic composition were performed before the analyses were done on the alpha emission probabilities, and on the neutron and photon sources as a result of (α,n) reaction. The main focus was on the quantitative evaluation of the neutron yield and the neutron energy spectrum for the generated neutrons through (α,n) reaction on light nuclei of dispersed nuclear fuel. Tests were carried out with the aim of creating an efficient calculation tool for the initial evaluation of the radiation characteristics for the irradiated multilayer nuclear fuel with different configurations and compositions.