Ultrahigh energy resolution gamma-ray spectrometers for precision measurements of uranium enrichment

Superconducting gamma-ray detectors offer an order of magnitude higher energy resolution than conventional high-purity germanium detectors. This can significantly increase the precision of non-destructive isotope analysis for nuclear samples where line overlap affects the errors of the measurement. We have developed gamma-detectors based on superconducting molybdenum-copper sensors and bulk tin absorbers for nuclear science and national security applications. They have, depending on design, an energy resolution between ∼50 and ∼150 eV FWHM at ∼100 keV. Here, we apply this detector technology to the measurement of uranium isotope ratios, and discuss the trade-offs between energy resolution and quantum efficiency involved in detector design.     

Design Features of a High Resolution Microcalorimeter EDX System

 High resolution, superconducting detectors allow energy dispersive X-ray spectrometry (EDX) with energy resolution and energy threshold far beyond the levels obtained with semiconductor detectors. These cryogenic detectors are run at temperatures of less than 100 mK and combine the excellent energy resolution of wavelength dispersive X-ray spectrometry (WDX) with the fast, energy dispersive analysis of EDX. CSP cryogenic spectrometer’s microcalorimeter type EDX cryodetectors are equipped with a mechanical cooling system that runs vibration free and allows completely automated operations on scanning electron microscopes (SEMs), field emission guns (FEGs) and transmission electron microscopes (TEMs). This detector type offers new opportunities in material analysis, especially when low excitation energies are applied or light elements are to be determined.     

Comparison of four types of gamma- and X-ray detectors for environmental applications in the 10–450 keV energy range

The performance of four types of - and X-ray radiation detectors for environmental applications was evaluated in the 10–450 keV energy range. Two cadmium zinc telluride (CdZnTe) room temperature semiconductor detectors were evaluated along with a cryogenically cooled semiconductor detector and two different types of scintillation detectors. The energy resolution, absolute peak detection efficiency and peak-to-background ratio of each of the detectors were calculated and intercompared. The advantages and disadvantages of environmental applications of each detector, along with their performance results, are summarized.     

Synchrotron radiation FTIR imaging in minutes: a first step towards real-time cell imaging

FTIR microscopy with a focal plane array (FPA) of detectors enables routine chemical imaging on individual cells in only a few minutes. The brilliance of synchrotron radiation (SR) IR sources may enhance the signal obtained from such small biosamples containing small amounts of organic matter. We investigated individual cells obtained from a cell culture specifically developed for transmission FTIR imaging using either a Globar or an SR source coupled to the same instrumentation. SR-IR source focussing was optimized to control the energy distribution on the FPA of detectors. Here we show that accessing the IR absorption distribution from all the organic contents of cells at 1 × 1 μm pixel resolution was possible only with high circulating current (≥1.2 A) illuminating a limited number of the FPA’s detectors to increase the signal-to-noise ratio of IR images. Finally, a high-current SR ring is mandatory for collecting FTIR images of biosamples with a high contrast in minutes.     

核化学与放射化学的研究进展

在我国核能快速发展的新形势下,新型核能资源的开发、乏燃料后处理、放射性废物处理与处置等核燃料循环化学研究日益活跃。随着科学技术的不断发展,离子加速器、反应堆、各种类型的探测器和分析设备、以及计算机技术等的发展,核化学与放射化学研究的范围和成果在不断扩展和增加,如核安全、环境放射化学、放射分析化学、放射性药物与标记化合物等,研究成果对于国防建设、核能发展、核技术应用等方面具有重要支撑作用。本文综述了近年来国内在上述领域所取得的研究进展。共引用参考文献161篇。     

Integrated analysis of video and radiation data for nuclear material monitoring system

The analysis algorithm, which simultaneously interprets the transfer route and types of nuclear materials and the types of transfer container, has been developed for the radioactive material transfers in the nuclear facility. This method can be accomplished by integration of video images based on pixel differences and radiation data imported by NDA radiation sensors in real time. This technology, based on pattem recognition by neural networks, is well suited for surveillance systems of large automated facilities, spent fuel storage facilities and new conceptual hot cell facilities such as DUPIC(Direct Use of spent PWR fuel in CANDU reactors) facility.     

Isotopic concentration of uranium from alpha spectrum of electrodeposited source on 4H-SiC detector at 500 °C

4H-SiC alpha detectors were fabricated with a 21-μm thick depletion depth and were packaged into a stainless-steel casing with a mineral insulation cable and a standard BNC connector. The packaged detectors had a resolution of 0.624% FWHM at 5.486 MeV prior to salt immersion. The detectors were then immersed in a LiCl–KCl–UCl₃ molten salt at 500 °C, from which a thin layer of depleted uranium was electrodeposited onto the detectors. Alpha particle emission spectra were collected from the electrodeposited source. The energy resolution of the surviving detector was 2.29% FWHM at 4.198 MeV and was sufficient to separate the ²³⁴U from ²³⁸U alpha emissions (577 keV difference). The ²³⁴U/²³⁸U activity ratio and the isotopic concentrations of ²³⁴U and ²³⁸U were determined and are representative of the uranium source used in the electrodeposition.

     

Development of high efficiency, multi-element CdZnTe detectors for portable measurement applications

We describe the development of detector arrays and electronics for large-volume, hand-held CdZnTe detectors with the same counting efficiency as portable NaI(Tl) detectors presently used for nuclear material measurement applications. The pulse-height resolution of the multi-element detectors is at least three times better than NaI(Tl) over a wide energy range (from 100 keV to several MeV), enabling more accurate measurements of gamma-rays emitted by special nuclear material. Arrays of up to eight coplanar grid detectors can be combined to make detectors ranging in size from 4 to 14 cm³. Because the number of spectroscopy channels is small, low-power, hand-held detectors can be manufactured with conventional printed circuit board technology, thus keeping the cost of multi-element detectors to a minimum. The design and performance of an 8-element detector is presented.     

Identification of unknown irradiated nuclear fuel through its fission product content

A procedure is demonstrated, through a simulation study, for the determination of the origin of unknown spent nuclear fuel, an important and challenging task in nuclear forensics. The procedure is an isotopic fingerprinting method relying on the fission product content of the unknown. The ‘unknown’ nuclear material is represented by the spent nuclear fuel of known origin in order to demonstrate the method and verify its predictive capabilities. The method is based on the comparison of the fission product compositions of the ‘unknown’ material and simulated known spent fuels from a range of commercial nuclear power stations using the multivariate statistical technique of factor analysis. Then, the provenance of the ‘unknown’ spent fuel is the commercial fuel with which it exhibits the highest similarity with respect to the fission product content.     

Performance of CdTe detector in the 13–1333 keV energy range

The photopeak efficiency, peak to valley ratio and energy resolution of a 3×3×1 mm³ CdTe detector were determined experimentally for 13–1333 keV photon energy by using polyester coated radioisotopes ²⁴¹Am, ²²Na, ⁵⁴Mn, ⁵⁷Co, ⁶⁰Co, ¹⁰⁹Cd, ¹³⁷Cs and ¹³³Ba. The data were analyzed by using PX4 from Amptek. The experimental values were fitted to an analytical function of photon energy, and an agreement was observed for the entire range of the studied energies. The results have shown that the CdTe detector has a high performance due to both the improved charge collection efficiency comparable with that of SI GaAs detectors (Vittone et al., 1999), and the low leakage current. Also, CdTe detector is very attractive for field application as it works at room temperature.     

Effect of gamma irradiation on the etching and optical properties of a newly developed nuclear track detector called (PNADAC) homopolymer

In the present work, we have determined the bulk-etch rates of a newly developed track detector called poly-[N-allyloxycarbonyl diethanolamine-bis allylcarbonate] (PNADAC) homopolymer at different temperatures to deduce its activation energy. The energy of activation is found to be (1.02±0.04) eV. This compares very well with the values of activation energy reported in the literature for the most commonly used nuclear track detectors. The effects of gamma irradiation on this new detector in the dose range of 4.7–14.5 Mrad have also been studied using UV–visible spectroscopic technique. The optical band gaps of the unirradiated and the gamma-irradiated detectors determined from the UV–visible spectra were found to decrease with the increase in gamma dose. These results have been explained on the basis of scission of the detector due to gamma irradiation.     

Fast evaluation of <Superscript>235</Superscript>U/<Superscript>238</Superscript>U ratio in nuclear spent fuel safe guard by thermal ionization mass spectrometry by using refractory metal oxides

A direct simple and fast method was established, to overcome the influence of low and high level impurities on the measurement of ²³⁵U/²³⁸U isotopic ratio in nuclear spent fuel safeguard by thermal ionization mass spectrometry (TIMS), by using refractory metal oxide. The addition of refractory metal oxides forming solution (RMOFS), in certain proportions alongside with the spent fuel solution on the sample filaments were found to be useful during the analysis of uranium isotopic ratio by TIMS. RMOFS (with oxide melting point exceeding 2,000 °C), and particularly that of magnesium, were found to be very effective in improving the quality of the ion signal of ²³⁵U and ²³⁸U, when added without the need for prior purification. Solutions of chromium, cerium, thorium, and magnesium were investigated, to select the more convenient one, and it was found that magnesium was very useful to start with. The method was very simple, improve both the accuracy and precision of the collected data, reduce the time required to achieve steady uranium pilot signal, and hence the over all time of the analysis, regardless of the level of impurities present.     

Extraction of neodymium from other fission products by co‐reduction of Sn and Nd

High temperature processing is an important method for recovering long‐lived elements from spent nuclear fuel. Electrolysis is the key technology for high temperature processing. The electrochemical behaviors of Sn²⁺, Nd³⁺ and the mechanisms of Sn‐Nd alloy formation were investigated on a Mo electrode at 873 K by conducting a series of electrochemical techniques. The results showed the deposition of Nd on inert electrode is a two‐step process in LiCl‐KCl‐SnCl₂ (2.0 wt.%) melt system. Subsequently, the electrochemical extraction of Nd from molten chlorides were carried out on the Mo electrode at temperature of 873 K by the potentiostatic electrolysis at ?1.2 V for 40 hr. Besides, the extraction efficiency is 97.6%. A series of potentiostatic electrolysis were carried out at potential range between ?1.0 and ? 1.4 V. The NdSn₃ alloy was obtained by electrolysis at ?1.2 V. This deposition potential is consistent with the predicted results of the mathematical model. The micro‐chemical analysis and morphology analysis of the deposits was characterized by energy dispersive spectrometry (EDS) with scanning electron microscopy (SEM) equipped. The composition of the deposits was analyzed by X‐ray diffraction (XRD) and inductive coupled plasma atomic emission spectrometer (ICP‐AES).     

Laser-induced breakdown spectroscopy—From research to industry,new frontiers for process control

This paper presents R&D activities to explore new laser parameter ranges in pulse energy, time and space for laser-induced breakdown spectroscopy. The collinear double pulse effect, which is well studied for pulses of typically several 100 mJ energy can also be observed for laser pulses having a pulse energy two orders of magnitude lower. In this case, maximum line emission intensity occurs at interpulse separations of a few 100 ns. Temporal pulse tailoring to improve the performance of LIBS is only a first step. A comprehensive approach includes spatial pulse shaping to generate craters with predefined shape or to improve spatial averaging for the analysis of inhomogeneous samples. High performance components for LIBS systems such as spectrometers, electronics and sample stands are required to enable industrial applications. Latest developments offer wide-band single spectra acquisition with a high spectral resolution at a measuring frequency of up to 500 Hz. The next generation of multi-channel integrator electronics for Paschen–Runge spectrometers equipped with PMT detectors will further push the measuring speed to up to 5 kHz, thus opening a new area of high-speed LIBS microanalysis. Novel LIBS devices for various industrial applications presented include analysis of metallic process control samples with scale layers, on-site analysis of slag samples in secondary metallurgy, high-speed identification of Al scrap, mix-up detection of pipe fittings as well as recent work towards in-process identification of hot coils in a rolling mill.     

Determination of 90Sr / 238U ratio by double isotope dilution inductively coupled plasma mass spectrometer with multiple collection in spent nuclear fuel samples with in situ 90Sr / 90Zr separation in a collision-reaction cell

Strontium-90 is one of the most important fission products generated in nuclear industry. In the research field concerning nuclear waste disposal in deep geological environment, it is necessary to quantify accurately and precisely its concentration (or the ⁹⁰Sr / ²³⁸U atomic ratio) in irradiated fuels. To obtain accurate analysis of radioactive ⁹⁰Sr, mass spectrometry associated with isotope dilution is the most appropriated method. But, in nuclear fuel samples the interference with ⁹⁰Zr must be previously eliminated. An inductively coupled plasma mass spectrometer with multiple collection, equipped with an hexapole collision cell, has been used to eliminate the ⁹⁰Sr / ⁹⁰Zr interference by addition of oxygen in the collision cell as a reactant gas. Zr⁺ ions are converted into ZrO⁺, whereas Sr⁺ ions are not reactive.A mixed solution, prepared from a solution of enriched ⁸⁴Sr and a solution of enriched ²³⁵U was then used to quantify the ⁹⁰Sr / ²³⁸U ratio in spent fuel sample solutions using the double isotope dilution method. This paper shows the results, the reproducibility and the uncertainties that can be obtained with this method to quantify the ⁹⁰Sr / ²³⁸U atomic ratio in an UOX (uranium oxide) and a MOX (mixed oxide) spent fuel samples using the collision cell of an inductively coupled plasma mass spectrometer with multiple collection to perform the ⁹⁰Sr / ⁹⁰Zr separation. A comparison with the results obtained by inductively coupled plasma mass spectrometer with multiple collection after a chemical separation of strontium from zirconium using a Sr spec resin (Eichrom) has been performed. Finally, to validate the analytical procedure developed, measurements of the same samples have been performed by thermal ionization mass spectrometry, used as an independent technique, after chemical separation of Sr.     

Study of silicon detectors for high resolution radioxenon measurements

Measurement of radioactive xenon in the atmosphere is one of several techniques to detect nuclear weapons testing, typically using either scintillator based coincidence beta/gamma detectors or germanium based gamma only detectors. Silicon detectors have a number of potential advantages over these detectors (high resolution, low background, sensitive to photons and electrons) and are explored in this work as a possible alternative. Using energy resolutions from measurements and detection efficiencies from simulations of characteristic electron and photon energies, the minimum detectable concentration for Xe isotopes was estimated for several possible detector geometries. Test coincidence spectra were acquired with a prototype detector.     

First shrinkage parameters of Slovak bentonites considered for engineered barriers in the deep geological repository of high-level radioactive waste and spent nuclear fuel

The high potential of bentonites to volume changes depending on the water content is considered as their advantage for the engineered barriers in the deep geological repository of high-level radioactive waste and spent nuclear fuel because of swelling and self-healing of cracks in contact with water. On the other hand, drying may lead to opening of cracks and spaces between the bentonite blocks. This would increase the permeability and contamination risk around the hot container with high-level radioactive waste and spent nuclear fuel, especially if the host rock mass is dry. First shrinkage tests on four Slovak bentonites studied for engineered barriers were carried out. The water content at the shrinkage limit and the relative linear shrinkage are the first available shrinkage parameters received for the bentonite paste. The shrinkage hazard is higher in the best bentonites with high swelling potential—from Kopernica and Jel?ový potok. The results indicated the necessity of further shrinkage tests to determine the relative linear and volume shrinkage of bentonite elements pressed of the loose bentonite powder of low water content.     

Recovery of palladium from basic reprocessing waste of spent nuclear fuel

The methods including collection method, extraction-collection method, and special extraction-collection method have been investigated for high speed and efficient recovery of palladium from high pH reprocessing waste of spent nuclear fuel. The equilibrium of the reactions can be obtained is less than 1 minute. The maximum percent recovery of Pd is about 89%, 96% and 97% for collection, extraction-collection, and special extraction-collection methods, respectively. Nearly 100% of back extraction of Pd in the organic phase can be attained by using 7.4M ammonia solution, with a phase ratio of 1:1. The purity of the Pd product is high. The percent recovery of Pd is constant, up to 5·10³ Gy of irradiation dose.