Determination of 94Nb in radioactive waste using ion exchange chromatography

This paper describes a rapid method of ⁹⁴Nb pre-concentration, separation and purification by using cation and anion exchange resins. The method is suitable for analyzing highly contaminated radioactive waste samples in a relatively short time and high decontamination factors. The use and effectiveness of the method was successfully tested by analysis of samples from nuclear reactor parts such as control rod drive shaft, shielding cassettes, neutron in-core measurement channels (KNI), pressure vessel construction material and fuel cassette construction material samples.     

Replacement of lead bricks with non-hazardous polymer-bismuth for low-energy gamma shielding

Lead is a hazardous substance, making it a disposal and industrial hygiene problem. The potential for creating mixed waste or mixed TRU waste exists if the lead becomes contaminated. The disposal of either waste stream is extremely difficult and costly. Bismuth is a non-hazardous material with shielding characteristics similar to lead. An HPGe was characterized using detector shielding composed of lead and polyethylene-based-bismuth to compare the shielding efficacy of both materials. Polymer-bismuth bricks may be recommended as gamma spectrometer shielding for sensitive low-energy measurements.     

Digital pulse-shape discrimination applied to an ultra-low-background gas-proportional counting system: first results

A new ultra-low-background proportional counter design was recently developed at Pacific Northwest National Laboratory (PNNL). This design, along with an ultra-low-background counting system which provides passive and active shielding with radon exclusion, has been developed to complement a new shallow underground laboratory (~30 m water-equivalent) constructed at PNNL. After these steps to mitigate dominant backgrounds (cosmic rays, external gamma-rays, radioactivity in materials), remaining background events do not exclusively arise from ionization of the proportional counter gas. Digital pulse-shape discrimination (PSD) is thus employed to further improve measurement sensitivity. In this work, a template shape is generated for each individual sample measurement of interest, a “self-calibrating” template. Differences in event topology can also cause differences in pulse shape. In this work, the temporal region analyzed for each event is refined to maximize background discrimination while avoiding unwanted sensitivity to event topology. This digital PSD method is applied to sample and background data, and initial measurement results from a biofuel methane sample are presented in the context of low-background measurements currently being developed.     

Method to assess the radionuclide inventory of irradiated graphite waste from gas-cooled reactors

About 17,000 tons of irradiated graphite waste will be produced from the decommissioning of the six gas-cooled nuclear reactors operated by Electricité De France Limited company (EDF), an energetic utility (http://www.edf.com/the-edf-group-42667.html) in France. Determining the radionuclide content of this waste is an important legal commitment for both safety reasons and in order to determine the best suited management strategy. As evidenced by numerous studies nuclear graphite is a very pure material, however, it cannot be considered from an analytical viewpoint as a usual homogeneous material. Radionuclide measurements in irradiated graphite exhibit very high discrepancies especially when corresponding to precursors at trace level. This huge discrepancy cannot be avoided and can be easily explained by Pierre Gy’s theory of sampling of finely divided materials. The assessment of a radionuclide inventory only based on few number of radiochemical measurements leads in most of cases to a gross over or under-estimation that can be detrimental to graphite waste management. A method using an identification calculation–measurement process is proposed in order to assess the radionuclide inventory as precisely as possible whilst guaranteeing an upper margin corresponding to a 2.5 % risk of under-assessment. This method closely reflects the reality of the main phenomenon at radionuclide origin in a reactor, while also incorporating the secondary effects that can alter this result such as radionuclide or precursor release during reactor operation.     

Neutron Equivalent Dose Rate Measurements of Gypsum-Waste Tire Rubber Layered Structures

The neutron equivalent dose rates (µSv/h) of gypsum, steel-reinforced rubber waste tire, and gypsum-waste tire rubber sandwich composite samples were investigated. Prepared samples were irradiated with ²⁴¹Am-Be neutrons and transmission values were obtained using dose equivalent rates measured with a BF₃ neutron detector. Results were compared to those of concrete, and as a result of neutron shielding, the performance of gypsum, waste tire, and waste tire (steel-reinforced rubber) embedded gypsum samples was higher than that of concrete. This information may be useful for shielding design of nuclear application areas.     

Short-time activation analysis with and without A6LiD converter

Neutron activation analysis (NAA) at various irradiation and measurement systems of the Institute's TRIGA mark II reactor provides a choice of sample oriented analyzing systems. This is demonstrated on the analysis of waste products, on an analysis of the phase content of a high-T_c superconductor and on a fluorine analysis in a standard reference material.     

Photofission for the detection of actinides

Nuclear waste management cost at CEA is a heavy budget share. Therefore, it is important, in order to minimize that cost, to select the most suitable repository for the waste. Significant cost savings are obtained, for example, whenever a medium activity waste drum initially filed to end up in a geological repository can be measured accurately enough to be accepted by the Centre de Stockage de l’Aube (CSA). Photofission can be used as a non-destructive technique to measure the actinide content embedded in a nuclear waste drum. It can even be applied when the waste is packaged in its heavy concrete shielding. It is equally sensitive to non-fissile nuclear material. However, the present state of art suffers from a lack of sensitivity to be able to help redirecting the waste to the proper repository. A new program called INPHO is proposed at CEA Saclay aiming at enhancing the sensitivity sufficiently to be applicable for waste characterization.     

Certified reference materials for the quality control of analysis in the environment

Summary The Community Bureau of Reference (BCR) has undertaken the production of a wide range of certified reference materials (CRMs) for the improvement of accuracy and measurement harmony in various areas of concern at European Community level. This paper provides an overview of the sectors of the BCR-programme in the field of environment i.e. for water systems (water, sediment, biota), soil analysis (soil, fertiliser, waste), plant analysis (aquatic and terrestrial plants), analysis related to animals (e.g. muscle or liver) or man (e.g. industry and food).     

Construction of a shallow underground low-background detector for a CTBT radionuclide laboratory

The International Monitoring System is a verification component of the Comprehensive Nuclear-Test-Ban Treaty, and in addition to a series of radionuclide monitoring stations, contains 16 radionuclide laboratories capable of verification of radionuclide station measurements. This paper presents an overview of a new commercially obtained low-background detector system for radionuclide aerosol measurements recently installed in a shallow (>30 meters water equivalent) underground clean-room facility at Pacific Northwest National Laboratory. Specifics such as low-background shielding materials and active shielding methods will be covered.     

A neutron monitor for in-situ real-time determination of transuranics in a processed waste effluent stream

A pilot plant operation at the Savannah River Site will remove ⁹⁰Sr, ¹³⁷Cs, and transuranics from a high-level liquid waste stream prior to encapsulation in a Saltstone Facility. Monitors are required to determine the concentrations of all radionuclides, including transuranics, in real-time on this processed waste stream. A neutron counter used to measure the concentration of each actinide isotope present in the stream is described. The neutron counter assembly consists of nested annular layers of shielding, reflectors, detectors, and moderators. On-line, live-time system control and calibration is provided by a time-tagged neutron source embedded in the moderator assembly.     

Study of the gamma spectrum of 16N with a BGO detector, for the purpose of calibration and of determining the fluorine grade of mineral samples

Over the past several years, the Pacific Northwest National Laboratory (PNNL) has developed an ultra-low-background proportional counter (ULBPC) technology. The resulting detector is the product of an effort to produce a low-background, physically robust gas proportional counter for applications like radon emanation measurements, groundwater tritium, and ³⁷Ar. In order to fully take advantage of the inherent low-background properties designed into the ULBPC, a comparably low-background dedicated counting system is required. An ultra-low-background counting system (ULBCS) was recently built in the new shallow underground laboratory at PNNL. With a design depth of 30 m water-equivalent, the shallow underground laboratory provides approximately 100× fewer fast neutrons and 6× fewer muons than a surface location. The ULBCS itself provides additional shielding in the form of active anti-cosmic veto (via 2-in-thick plastic scintillator paddles) and passive borated poly (1 in.), lead (6 in.), and copper (~3 in.) shielding. This work will provide details on PNNL’s new shallow underground laboratory, examine the motivation for the design of the counting system, and provide results from the characterization of the ULBCS, including initial detector background.     

Radiochemical methods in nuclear waste management

Radiochemical techniques play an important role in nuclear waste management studies at the Centre for Analytical Research in the Environment. Current research includes the measurement of atmospheric emissions from a reprocessing plant, modelling of environmental pathways, and evaluation of reactor materials for the prediction of releases from ultimate disposal sites. A radiochemical separation procedure has been developed for¹²⁹I, which has been applied to a study of emissions from the Sellafield nuclear reprocessing plant. Atmospheric emissions, rainfall, grass, milk and wild food are analysed to model the ultimate fate of¹²⁹I. A new procedure has been developed for measuring sub-ppm concentrations of chlorine in archive steel samples as part of programme to produce an inventory of arisings of³⁶Cl from nuclear power stations. The method has been extended to other reactor materials and this paper describes the application of the radiochemical method for Magnox alloys, mild steel and cast iron.     

Rapid method for determination of plutonium, americium and curium in large soil samples

The analysis of actinides in environmental soil and sediment samples is very important for environmental monitoring. There is a need to measure actinide isotopes with very low detection limits. A new, rapid actinide separation method has been developed and implemented that allows the measurement of plutonium, americium and curium isotopes in large soil samples (100–200 g) with high chemical recoveries and effective removal of matrix interferences. This method uses stacked TEVA Resin®, TRU Resin® and DGA-Resin® cartridges from Eichrom Technologies (Darien, IL, USA) that allows the rapid separation of plutonium (Pu), americium (Am), and curium (Cm) using a single multi-stage column combined with alpha-spectrometry. The method combines an acid leach step and innovative matrix removal using cerium fluoride precipitation to remove the difficult soil matrix. This method is unique in that it provides high tracer recoveries and effective removal of interferences with small extraction chromatography columns instead of large ion-exchange resin columns that generate large amounts of acid waste. By using vacuum box cartridge technology with rapid flow rates, sample preparation time is minimized.     

A comparison of quantum chemical models for calculating NMR shielding parameters in peptides: mixed basis set and ONIOM methods combined with a complete basis set extrapolation

This article compares several quantum mechanical approaches to the computation of chemical shielding tensors in peptide fragments. First, we describe the effects of basis set quality up to the complete basis set (CBS) limit and level of theory (HF, MP2, and DFT) for four different atoms in trans N-methylacetamide. For both isotropic shielding and shielding anisotropy, the MP2 results in the CBS limit show the best agreement with experiment. The HF values show quite a different tendency to MP2, and even in the CBS limit they are far from experiment for not only the isotropic shielding of carbonyl carbon but also most shielding anisotropies. In most cases, the DFT values differ systematically from MP2, and small basis-set (double- or triple-zeta) results are often fortuitously in better agreement with the experiment than the CBS ones. Second, we compare the mixed basis set and ONIOM methods, combined with CBS extrapolation, for chemical shielding calculations at a DFT level using various model peptides. From the results, it is shown that the mixed basis set method provides better results than ONIOM, compared to CBS calculations using the nonpartitioned full systems. The information studied here will be useful in guiding the selection of proper quantum chemical models, which are in a tradeoff between accuracy and cost, for shielding studies of peptides and proteins.     

A new system of gamma-ray spectrometry for activation analysis

A new system of γ-ray spectrometry has been designed and constructed. This system performs a precise, quick and automatic activation analysis using our nuclear reactor (TRIGA-II Research Reactor) as a neutron source. The main components are detector system consisted of a high resolution Ge(Li) detector settled in a shielding cabinet and a well-type NaI(Tl) scintillation counter for anti-Compton measurement, and data acquisition and spectra analysis system performed independently by two mini computer. The system has operated for atmospheric aerosols as an example of activation analysis.     

A novel shielding material prepared from solid waste containing lead for gamma ray

Human beings are continuously exposed to cosmogenic radiation and its products in the atmosphere from naturally occurring radioactive materials (NORM) within Earth, their bodies, houses and foods. Especially, for the radiation protection environments where high ionizing radiation levels appear should be shielded. Generally, different materials are used for the radiation shielding in different areas and for different situations. In this study, a novel shielding material produced by a metallurgical solid waste containing lead was analyzed as shielding material for gamma radiation. The photon total mass attenuation coefficients (μ/ρ) were measured and calculated using WinXCom computer code for the novel shielding material, concrete and lead. Theoretical and experimental values of total mass attenuation coefficient of the each studied sample were compared. Consequently, a new shielding material prepared from the solid waste containing lead could be preferred for buildings as shielding materials against gamma radiation.     

Shielding design and detection of neutrons from medical and industrial electron accelerators--simple method of design calculation for neutron shielding

The neutron leakage from medical and industrial electron accelerators has become an important problem and its detection and shielding is being performed in their facilities. This study provides a new simple method of design calculation for neutron shielding of those electron accelerator facilities by dividing into the following five categories; neutron dose distribution in the accelerator room, neutron attenuation through the wall and the door in the accelerator room, neutron and secondary photon dose distributions in the maze, neutron and secondary photon attenuation through the door at the end of the maze, neutron leakage outside the facility-skyshine.     

Determination of electron density in an atomic plasma by least-squares fit to the Stark profile

The measurement of electron density by a unique, one-parameter least-squares fit to the Stark-broadened hydrogen emission line is presented. The curve-fit approach utilizes approximately 60–80 points, and provides improved precision in the measurement.A comparison of data obtained by measurement of the width at half-height and by least-squares analysis indicates that the half-width method exhibits substantial systematic errors, generally over-estimating electron density by a factor of 2–3.