The matrix method for radiological characterization of radioactive waste

Beam losses are responsible for material activation in some of the components of particle accelerators. The activation is caused by several nuclear processes and varies with the irradiation history and the characteristics of the material (namely chemical composition and size). Once at the end of their operational lifetime, these materials require radiological characterization.The radionuclide inventory depends on the particle spectrum, the irradiation history and the chemical composition of the material. As long as these factors are known and the material cross-sections are available, the induced radioactivity can be calculated analytically. However, these factors vary widely among different items of waste and sometimes they are only partially known.The European Laboratory for Particle Physics (CERN, Geneva) has been operating accelerators for high-energy physics for 50 years. Different methods for the evaluation of the radionuclide inventory are currently under investigation at CERN, including the so-called “matrix method”. This paper provides a mathematical formulation of the matrix method highlighting its advantages and limits of validity.     

Radiation effects on polymers for coatings on copper canisters used for the containment of radioactive materials

The present work proposes applying polyurethane coatings as an additional barrier in the design of Canadian nuclear waste disposal containers. The goal of the present research is to investigate the physico-mechanical integrity of a natural castor oil-based polyurethane (COPU) to be used as a coating material in pH-radiation-temperature environments. As the first part to these inquiries, the present paper investigates the effect of a mixed radiation field supplied by a SLOWPOKE-2 nuclear research reactor on COPUs that differ only by their isocyanate structure. FTIR, DSC, DMA, WAXS, and MALDI are used to characterize the changes that occur as a result of radiation and to relate these changes to polymer structure and composition. The COPUs used in the present work have demonstrated sustained physico-mechanical properties up to accumulated doses of 2.0 MGy and are therefore suitable for end-uses in radiation environments such as those expected in the deep geological repository.     

Application of a chronoamperometric measurement to the on-line monitoring of a lithium metal reduction for uranium oxide

Both a potentiometric and a chronoamperometric electrochemical technique have been applied in an attempt to develop an efficient method for an on-line monitoring of a lithium metal reduction process of uranium oxides at a high-temperature in a molten salt medium. As a result of this study, it was concluded that the chronoamperometric method provided a simple and effective way for a direct on-line monitoring measurement of a lithium metal reduction process of uranium oxides at 650 °C by the measuring electrical currents dependency on a variation of the reduction time for the reaction. A potentiometric method, by adopting a homemade oxide ion selective electrode made of ZrO₂ stabilized by a Y₂O₃ doping, however, was found to be inappropriate for an on-line monitoring of the reduction reaction of uranium oxide in the presence of lithium metal due to an abnormal behavior of the adopted electrodes. The observed experimental results were discussed in detail by comparing them with previously published experimental data.     

Determination of Fe in waste waters of the Krsko nuclear power plant measured simultaneously by liquid scintillation spectrometer (LSC) and X-ray spectrometer (XRS)

This paper presents comparison of two methods for the determination of ⁵⁵Fe activity of waste waters discharged from the Krsko nuclear power plant (KNPP). Research was conducted on 12 composite samples of waste water collected in the waste monitor tank (WMT) during each month as well as on Analytics, Inc. cross-check sample. Results showed that the complicated and time-consuming method proposed by the Environmental Measurements Laboratory (EML) could be successfully replaced with a simple and fast based on the extraction of ⁵⁵Fe from waste water by non-specific chelating agent ammonium-pyrrolidinedithiocarbamate (APDC) at pH 4 after separation from cobalt, and activity measurement by X-ray fluorescence spectroscopy (XRS). Results obtained by the XRS method were approximately 8.6% lower than those obtained by liquid scintillation spectrometer (LSC). The mean deviation of the XRS results from the activity of cross-check sample was 2.47%, which ensures that this method is accurate enough for environmental monitoring.     

Fusion-Fission Transmutation Scheme—Efficient destruction of nuclear waste

A fusion-assisted transmutation system for the destruction of transuranic nuclear waste is developed by combining a subcritical fusion-fission hybrid assembly uniquely equipped to burn the worst thermal nonfissile transuranic isotopes with a new fuel cycle that uses cheaper light water reactors for most of the transmutation. The center piece of this fuel cycle, the high power density compact fusion neutron source (100 MW, outer radius <3 m), is made possible by a new divertor with a heat-handling capacity five times that of the standard alternative. The number of hybrids needed to destroy a given amount of waste is an order of magnitude below the corresponding number of critical fast-spectrum reactors (FRs) as the latter cannot fully exploit the new fuel cycle. Also, the time needed for 99% transuranic waste destruction reduces from centuries (with FR) to decades.     

Improvement of non-destructive fissile mass assays in α low-level waste drums: A matrix correction method based on neutron capture gamma-rays and a neutron generator

Within the framework of radioactive waste control, non-destructive assay (NDA) methods may be employed. The active neutron interrogation (ANI) method is now well-known and effective in quantifying low α-activity fissile masses (mainly ²³⁵U, ²³⁹Pu, ²⁴¹Pu) with low densities, i.e. less than about 0.4, in radioactive waste drums of volumes up to 200 l. The PROMpt Epithermal and THErmal interrogation Experiment (PROMETHEE [F. Jallu, A. Mariani, C. Passard, A.-C. Raoux, H. Toubon, Alpha low level waste control: improvement of the PROMETHEE 6 assay system performances. Nucl. Technol. 153 (January) (2006); C. Passard, A. Mariani, F. Jallu, J. Romeyer-Dherber, H. Recroix, M. Rodriguez, J. Loridon, C. Denis, PROMETHEE: an alpha low level waste assay system using passive and active neutron measurement methods. Nucl. Technol. 140 (December) (2002) 303-314]) based on ANI has been under development since 1996 to reach the incinerating α low level waste (LLW) criterion of about 50 Bq[α] per gram of crude waste (≈50 μg Pu) in 118 l drums on the date the drums are conditioned.Difficulties arise when dealing with matrices containing neutron energy moderators such as H and neutron absorbents such as Cl. These components may have a great influence on the fissile mass deduced from the neutron signal measured by ANI. For example, the calibration coefficient measured in a 118 l drum containing a cellulose matrix (density d = 0.144 g cm⁻³) may be 50 times higher than that obtained in a poly-vinyl-chloride matrix (d = 0.253 g cm⁻³). Without any information on the matrix, the fissile mass is often overestimated due to safety procedures and by considering the most disadvantageous calibration coefficient corresponding to the most absorbing and moderating calibration matrix.The work discussed in this paper was performed at the CEA Nuclear Measurement Laboratory in France. It concerns the development of a matrix effect correction method, which consists in identifying and quantifying the matrix components by using prompt gamma-rays following neutron capture. The method aims to refine the value of the adequate calibration coefficient used for ANI analysis.This paper presents the final results obtained for 118 l waste drums with low α-activity and low density. This paper discusses the experimental and modelling studies and describes the development of correction abacuses based on gamma-ray spectrometry signals.     

Spark Plasma Sintering of simulated radioisotope materials within tungsten cermets

A Spark Plasma Sintering (SPS) furnace was used to produce ceramic-metallic sinters (cermets) containing a simulated loading of radioisotope materials. CeO₂ was used to simulate loadings of PuO₂, UO₂ or AmO₂ within tungsten-based cermets due to the similar kinetic properties of these materials, in particular the respective melting points and Gibbs free energies. The work presented demonstrates the capability and suitability of the SPS process for the production of radioisotope encapsulates for nuclear fuels and other applications (including waste disposal and radioisotope power and heat source fabrication) where the mechanical capture of radioisotope materials is required.     

高放废物地质处置黏土岩处置库围岩研究现状

世界上很多国家都对处置库的可能围岩进行了详细研究。通过对比,认为花岗岩、黏土岩、岩盐比较适合作为处置库围岩,而黏土岩由于具有自封闭性、渗透率低等其他岩石类型不可比拟的优点,因而将黏土岩作为高放废物地质处置库围岩越来越受到各国的关注。文章同时介绍了瑞士、法国、比利时等国家在黏土岩中所进行的大量研究,均认为在黏土岩中处置高放废物和乏燃料是安全的。文章还对黏土岩处置库概念设计、黏土岩处置库围岩地下实验室研究,以及我国开展黏土岩处置库研究的意义等进行了综述。     

Release of radioactive materials from high active liquid waste in small-scale hot test for boiling accident in reprocessing plant

The release behavior of radioactive materials from high active liquid waste (HALW) has been experimentally investigated under boiling accident conditions. In the experiments using HALW obtained through laboratory-scale reprocessing, the release ratio was measured for fission product (FP) nuclides such as Ru, Tc-99, Cs, Sr, Nd, Y, Mo, Rh and actinides such as Cm-242 and Am-241. As a result, the release ratio was 0.20 for Ru and was around 1×10^?4 for the FP and actinide nuclides. Ru was released into the gas phase in the form of both mist and gas. For its released amount, weak dependency was found to its initial concentration in the test solution. The release ratio decreased with the increase in the initial concentration. For other FP nuclides and actinides as non-volatile, released into the gas phase in the form of mist, the released amount increased with the increase in the initial concentration. The release ratio of Ru and NOx concentration increased with the increase in the temperature of the test solutions. They were released together almost at the same temperature between 200 and 300 °C. Size distribution of particles like mist was measured. The data show that there was a difference between distributions at the temperatures below 150 °C and over 200 °C.     

System for high-level radioactive waste using microchannel chip — extraction behavior of metal ions from aqueous phase to organic phase in microchannel

In order to clarify the extraction behavior of U(VI) from aqueous phase to organic one in microchannel, we have carried out extraction experiments of U(VI) from HNO₃ aqueous solution of 3 M (M = mol/dm³) to 30% or 100% TBP phase in microchannel. From the results of extraction experiments, it was found that the extraction of U(VI) in microchannel could be performed in a short time for approximately 1 s with a good extractability in both organic phases of 30% and 100% TBP, and suggested that the other nuclides could be extracted with high extraction efficiency in microchannel. Furthermore, it is expected that the innovative and sophisticated nuclide separation systems can be developed by using microchannel extraction with selective extractants for specific nuclide.     

Thermal expansion of TRU nitride solid solutions as fuel materials for transmutation of minor actinides

The lattice thermal expansion of the transuranium nitride solid solutions was measured to investigate the composition dependence. The single-phase solid solution samples of (Np_0.55Am_0.45)N, (Pu_0.59Am_0.41)N, (Np_0.21Pu_0.52Am_0.22Cm_0.05)N and (Pu_0.21Am_0.18Zr_0.61)N were prepared by carbothermic nitridation of the respective transuranium dioxides and nitridation of Zr metal through hydride. The lattice parameters were measured by the high temperature X-ray diffraction method from room temperature up to 1478 K. The linear thermal expansion of each sample was determined as a function of temperature. The average thermal expansion coefficients over the temperature range of 293-1273 K for the solid solution samples were 10.1, 11.5, 10.8 and 8.8 × 10⁻⁶ K⁻¹, respectively. Comparison of these values with those for the constituent nitrides showed that the average thermal expansion coefficients of the solid solution samples could be approximated by the linear mixture rule within the error of 2-3%.