A new concept for safeguarding and labeling of long-term stored waste and its place in the scope of existing tagging techniques

The idea of a novel labeling method is suggested for a new way of long-term security identification, inventory tracking, prevention of falsification and theft of waste casks, copper canisters, spent fuel containers, mercury containers, waste packages and other items. The suggested concept is based on the use of a unique combination of radioisotopes with different predictable half life. As an option for applying the radioisotope tag to spent fuel safeguarding it is suggested to use a mixture of {\alpha}-emitting isotopes, such as 241Am etc., with materials that easily undergo {\alpha}-induced reactions with emission of specific {\gamma}-lines. Thus, the existing problem of the disposing of smoke detectors or other devices [1] which contain radioisotopes can be addressed, indirectly solving an existing waste problem. The results of the first pilot experiments with two general designs of storage canisters, namely a steel container which corresponds to the one which is commonly used for long-term storing of mercury in Europe and USA and a copper canister, the one which is in applications for nuclear repositories, are presented. As one of the options for a new labeling method it is proposed to use a multidimensional bar code symbology and tungsten plate with ultrasound techniques. It is shown that the new radioisotope label offers several advantages in the scope of existing tagging techniques (overview is given) and can be implemented even with low activity sources.Comment: Workshop - Scanning the Horizon: Novel Techniques and Methods for Safeguards, International Atomic Energy Agency, IAEA Headquarters in Vienna, Austria, 201

Derivation and quantitative analysis of the differential self-interrogation Feynman-alpha method

A stochastic theory for a branching process in a neutron population with two energy levels is used to assess the applicability of the differential self-interrogation Feynman-alpha method by numerically estimated reaction intensities from Monte Carlo simulations. More specifically, the variance to mean or Feynman-alpha formula is applied to investigate the appearing exponentials using the numerically obtained reaction intensities.Comment: Proceedings 52nd INMM conference, Palm Desert, 17-21 July 201

The neutron-gamma Feynman variance to mean approach: gamma detection and total neutron-gamma detection (theory and practice)

Two versions of the neutron-gamma variance to mean (Feynman-alpha method or Feynman-Y function) formula for either gamma detection only or total neutron-gamma detection, respectively, are derived and compared in this paper. The new formulas have a particular importance for detectors of either gamma photons or detectors sensitive to both neutron and gamma radiation. If applied to a plastic or liquid scintillation detector, the total neutron-gamma detection Feynman-Y expression corresponds to a situation where no discrimination is made between neutrons and gamma particles. The gamma variance to mean formulas are useful when a detector of only gamma radiation is used or when working with a combined neutron-gamma detector at high count rates. The theoretical derivation is based on the Chapman-Kolmogorov equation with inclusion of general reactions and passage intensities for neutrons and gammas, but with the inclusion of prompt reactions only. A one energy group approximation is considered. The comparison of the two different theories is made by using reaction intensities obtained in MCNPX simulations with a simplified geometry for two scintillation detectors and a 252Cf-source enclosed in a steel container. In addition, the variance to mean ratios, neutron, gamma and total neutron-gamma, are evaluated experimentally for a weak 252Cf neutron-gamma source in a steel container, a 137Cs random gamma source and a 22Na correlated gamma source. Due to the focus being on the possibility of using neutron-gamma variance to mean theories for both reactor and safeguards applications, we limited the present study to the general analytical expressions for Feynman-Y formulas

The Influence of Zn and Cd Accumulation on the Growth and Development of Medicinal Plants in the Impact Zone of the Novocherkassk Power Station

Over the pastdecade, particular attention has been paid to studies of the chemical composition of medical plants to identify the possible negative consequences of using raw plant material polluted with heavy metals for the production of medical drugs. In our study, we analyzed the chemical composition of the medical plants growing in the impact area of the Novocherkassk power station. Specifically, the plants Artemisia austriaca, Poa pratensis and Elytrigia repenswere examined for the analysis.The content and distribution of Zn and Cd, which are most distributed in industrial emissions and belong to the first class of hazardous elements, were measured. The maximum permissible content (MPC) of Zn in the raw material of Artemisia austriaca and Elytrigia repens was found, as was the maximum content of Cd in all analyzed plants growing in the 5km area around thepower station. The plant Artemisia austriacawasfound to have Zn and Cd accumulation in itsabovegroundcomponents, while in Poa pratensis and Elytrigia repens, accumulation was in the roots. The morphobiometric parameters of the plants were mostly dependent on the soil properties, followed by the degree of technogenic load. The content of Zn and Cd in the medical drugs was higher than the MPC without visible features of heavy metal pollution and so these plants weredangerous for human health. Keywords: heavy metals, technogenic load, phytoreagents, morphometric parameter

Application of the two-group - one-region and two-region - one-group Feynman-alpha formulas in safeguards and accelerator-driven system (ADS)

The applicability of the two-group (one-region) and two-region (one-group) Feynman-alpha (variance to mean) formulas was assessed with regards to applications in safeguards and accelerator-driven system (ADS) considered as an option for transmutation of nuclear wastes. Since two-group calculations with the master equation technique when both thermal and fast fissions are included, have not been performed earlier, investigation of this problem has a methodological value of its own. The potential applications of the two-group - one-region and two-region - one-group Feynman-alpha approaches in nuclear safeguards were evaluated and compared to the results of Monte-Carlo simulations

Time intervals matrix analysis of 235U and 239Pu content in a spent fuel assembly using Lead Slowing down Spectrometer

A key motivation for developing the technology capable of quantifying plutonium (Pu) in spent fuel assemblies with nondestructive assay (NDA) techniques is knowledge of the physical parameters of irradiated nuclear fuel which is important both for nuclear safeguards and nuclear fuel management. One of the most attractive NDA approaches applied for the determination of the total amount of plutonium using a pulsed neutron source is the method of slowing-down time spectrometry in lead where the energy spectrum of neutrons can be represented as being monoenergetic with minor deviation from the peak value in each time moment after a fast neutron pulse. This fact was successfully used in developing several methods of Pu mass determination and confirmed the potential of the Lead Slowing Down Spectrometer (LSDS) to get detailed information about spent fuel [1-2].A method which we presented earlier is based on a matrix of time intervals where large differences in the number of fissions of 235U and 239Pu are observed [3]. This technique allows increasing precision in the Pu evaluation by decreasing the self-shielding effect significantly. As opposed to homogeneous-volume approximations used in our previous research in this work we describe the detailed Monte Carlo models of real fuel assemblies, as well as the effects of the influence of the scintillation detector to the system in question. Although the proposed method for spent fuel assemblies characterization has only been studied using Monte Carlo simulations, it was possible to demonstrate the 239Pu determination using a DT pulsed neutron source, Lead Slowing Down Spectrometer and fast timing scintillatior which is sensitive to both photons and neutrons, and n-γ pulse shape discrimination allows to get additional information about the system