Isotopic Composition Of Plutonium Research Articles

Multivariate test applied to single Pu isotopic composition measurement

The determination of the plutonium isotopic composition by gamma spectrometry is one of the tasks performed by nuclear inspectors. It is typically measured using planar germanium detectors. The spectra are analysed with deconvolution codes, which calculate all isotopic abundances. The experimental results are then compared with the declared values and the acceptance criterion is often based on one unique isotope, 240Pu, to distinguish between reactor and weapon grade Pu.If the measurement is a part of a measurement series, it is possible to calculate covariance terms and carry out a more comprehensive comparison using Hotelling's test as recently presented in Berndt and Mortreau (2023) [1]. In this study, it was shown that the T2 test often rejects simulated false cases which are accepted if only the 240Pu abundance is considered. Hence, the T2 test should be preferred to the 240Pu criterion.In case of a single measurement, which is typical for inspections, the co-variance cannot be determined and the comprehensive calculation of the T2 test as performed in Berndt and Mortreau (2023) [1] cannot be applied. The purpose of the present paper is to consider approximations of the variance-covariance matrix to allow the application of the T2 test to isolated measurements. The results were then applied to more than 400 single measurements of Pu samples covering many different experimental conditions.

Evaluation of neutronic performance for the VVER-1000 reactor core with regenerated uranium-plutonium fuel

The possibility has been considered for the VVER-1000 reactor fuel loading to be formed based on regenerated fuel with the use of the spent fuel accumulated in reactors of the same type. A study was undertaken to investigate the change in the isotopic composition of the plutonium discharged from a thermal reactor in the course of its multiple reprocessing and recycle in a thermal neutron reactor. To obtain an equilibrium isotopic composition of the reactor-grade plutonium, 3D neutronic calculations were performed for the stationary fuel cycles of a VVER-1000 serial reactor with conventional oxide fuel and oxide fuel based on regenerated uranium and based on an undivided mixture of uranium and plutonium oxides from SNF. The neutronic performance of reactor cores was compared for the above mentioned fuel types in the course of the fuel company, including the following: in-core radial power density shaping, values of reactivity coefficients for various thermal parameters, reactivity control system efficiency, etc.

Effects of relative humidity, surface area and production route on hydrogen yields from water on the surface of plutonium dioxide

Over 60 samples of plutonium dioxide (PuO2) powders of varying provenance have been exposed to humid atmospheres and the hydrogen (H2) generation rates measured by gas chromatography. The effects of relative humidity (RH), specific surface area (SSA), plutonium isotopic composition (absorbed dose), and overlying atmosphere have been investigated for ‘as received’ PuO2 powders from the United Kingdom’s Thorp and Magnox reprocessing plants, high surface area powder produced in the laboratory and Magnox PuO2 that was recalcined at 800, 900 or 950°C. Hydrogen generation was shown to be susceptible to subtle influences with the most consistent results observed at 95% RH. However, it was shown that the measured (net) hydrogen decreases with decreasing RH and with atmosphere in the order: Air > argon > nitrogen. There was no clear effect of SSA, apart from with the highest SSA samples (∼40 m2.g−1) but these powders also have very different morphology, porosity, and carbon content to the rest. The results presented substantially enhance the growing body of literature on the factors that determine hydrogen and gas generation from PuO2 that has significant implications for long term safe storage of plutonium globally.

Hotelling’s test applied to measured isotopic compositions of plutonium

Hotelling’s test applied to measured isotopic compositions of plutonium

Distributions and impacts of plutonium in the environment originating from the Fukushima Daiichi Nuclear Power Plant accident: An overview of a decade of studies

This paper reviews the current knowledge on plutonium (Pu) isotopic composition (the atom or activity ratios) and activity concentrations of 238Pu, 239Pu, 240Pu, and 241Pu resulting from the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident in 2011. In this critical review, we document the characteristic values of Pu atom or activity ratios (fingerprints) and present their spatial distributions around the FDNPP site. Based on multiple Pu fingerprints (238Pu/239+240Pu activity ratio, 240Pu/239Pu atom ratio, and 241Pu/239Pu atom ratio), we clarify that Pu contamination from the FDNPP accident occurred in a restricted terrestrial area, while Pu in the Northwest Pacific Ocean is still predominately sourced from the Pacific Proving Grounds (PPG) and global fallout. Using a simple two end–member mixing model, we calculate average contributions of Pu from the FDNPP accident of 13 ± 20% (n = 180) in soil samples, 55 ± 32% (n = 38) in leaf litter samples, and 67 ± 26% (n = 129) in air dust/black substances. In the marine environment, the PPG source average contributions are 45 ± 15% (n = 76) in seawater and 42 ± 12% (n = 48) in sediments. The spatial distributions of Pu atom or activity ratios based on existing studies suggest that: 1) in the terrestrial region investigated 80 km northwest of the FDNPP site, the Pu contamination is mainly observed in an area within a 50 km distance, and 2) in the terrestrial region investigated 60 km southwest of the FDNPP site, the Pu contamination is mainly observed in an area within a 30 km distance. Studies of Cs-bearing radioactive particles indicate that Pu occurs as Pu oxide, and the fuel fragments containing Pu that were released from the reactors to the surrounding environment are associated with micron-scale Cs-bearing radioactive particles. We note that the fractionation between Pu and other radionuclides occurred after release. These new findings about the Pu fingerprints around the FDNPP site will help researchers to establish a reference background database for future environmental risk assessment and geochemical study there.

Simulating the fuel cycle of a lead-cooled fast reactor

The development of nuclear power with fast reactors is associated with the implementation of a closed nuclear fuel cycle (CNFC). In this regard, one actual task is to simulate the stages of the fuel cycle with study of the neutron-physical characteristics of the core. The design of a reactor for operation in the closed nuclear fuel cycle mode is impossible without the using of verified and certified software packages for calculating fast reactors, capable of simulating all stages of the operation of the reactor facility and the fuel cycle. For the calculations, the FACT-BR software package was used, which has all the necessary capabilities to simulate the operation of the reactor in the closed nuclear fuel cycle mode, taking into account the stages of fuel storage and refabrication. The article presents a technique for modeling the fuel cycle, implemented for the operation of fast reactors with a lead coolant. To demonstrate methodology, a closed nuclear fuel cycle was simulated for the BREST-OD-300 and BR-1200 reactors for the design life. The article describes the scenarios in which the calculation of the burnup of reactor was carried out. In the considered scenarios, it is assumed that the unloading of fuel at the end of the micro campaign is conducted according to the maximum burnup. During the computational modeling the ranges of changes in fuel density and enrichment, reactivity margin, breeding ratio and isotopic composition of plutonium were determined.

Uncertainty propagation in Pu isotopic composition calculation by gamma spectrometry: theory versus experiment

Abstract Calculation and reporting of total uncertainties are essential criteria in the nuclear industry since measured results are used in decision making. Plutonium isotopic compositions (Pu IC) in different matrices are required at various stages of a close-loop nuclear fuel cycle. Under- and/or over-estimating uncertainties in Pu IC may result in avoidable radiological emergencies. In this work, we present the uncertainty budget for Pu IC determination using 120–450 keV gamma emission lines recorded with an HPGe. Detailed uncertainty propagation equations based on the propagation of partial derivatives are constructed and solved. Effects of the individual uncertainties on the total uncertainty are studied for different counting durations starting from 5 min up to 24 h. Results are compared with TIMS results and the theoretically calculated uncertainties were verified with multiple experimental data.

Possible ways to solve the problems of using MOX fuel in thermal reactors

Modern nuclear power has a number of crucial problems, including the accumulation of spent nuclear fuel (SNF) and the projected shortage of nuclear resources. The conversion to a closed fuel cycle is the most promising solution to these problems. Mixed uranium-plutonium fuel (MOX fuel) has been successfully tested and commissioned under this concept by several European countries. However, its operation in reactors designed for uranium fuel has a number of problems, most of which have been resolved. Nevertheless, the spent MOX fuel has neutronic characteristics that do not allow its reuse in thermal reactors and solve the problem of SNF accumulation. This is due to the accumulation of heavy (even) plutonium isotopes, which are not fissionable in the thermal spectrum. In fast reactors, it is possible to create conditions under which the isotopic composition of plutonium from the spent MOX fuel from thermal reactors changes in such a way that it becomes suitable for reuse in thermal reactors. This process is called improving. The concept for the development of two-component nuclear energy based on the joint operation of thermal and fast neutron reactors and the creation of a closed fuel cycle was adopted in Russia to implement this process and solve crucial problems of nuclear power.

Irregularity of plutonium isotopic composition of the BREST-OD-300 initial load

This article is dedicated observation of irregularity of plutonium isotopic composition of the BREST-OD-300 initial load. The purpose of the research is to study the effect of the plutonium isotopic composition in the nitride fuel of the BREST–OD-300 reactor core starting load on the reactivity reserve. The calculations is creating by writing simplified three-dimensional geometry of the BREST-OD-300 reactor in software package “SERPENT”. There are five options in the fuel assemblies locations are considered with different cases of using the plutonium isotopic composition with deviations for some isotopes. Calculations have shown that the deviation of the plutonium isotope composition from the design values of the fuel loading of the BREST-OD-300 reactor does not affect the stability of reactivity during fuel burnout within the effective fraction of delayed neutrons. In the course of the research, changes in the effective multiplication coefficient are obtained over five years of reactor operation. The value of this indicator does not depend much on the plutonium deviation in the considered variations in the arrangement of fuel assemblies in the reactor core.

Plutonium in Southern Yellow Sea sediments and its implications for the quantification of oceanic-derived mercury and zinc

The spatial distributions of mercury (Hg) and zinc (Zn) concentration and the isotopic composition of plutonium (Pu) were investigated in surface sediments and sediment cores collected from the Southern Yellow Sea (SYS) during May 2014. The variation of the 240Pu/239Pu atom ratio (0.18–0.31) in the surface sediments of the SYS clearly indicated a signal of close-in fallout input from the Pacific Proving Ground (PPG). The buried 239+240Pu in the sediment of the SYS was estimated to be (4.7 ± 0.5) × 1010 Bq y−1 during the period from 2011 to 2014, of which ∼33% (1.5 × 1010 Bq y−1) was derived from the PPG by long-range transport via ocean currents (e.g., the North Equatorial Current and Kuroshio Current). The concentrations of Hg and Zn varied from 0.003 to 0.067 mg kg−1 and from 43.9 to 137 mg kg−1, respectively, and exhibited positive correlations with the 239+240Pu activity both in the surface sediments (0–1 cm) and upper layers (7 cm) of the sediment cores. Therefore, by using Pu as a tracer, we estimated that the oceanic input contributed 2.0 tons y−1 of Hg and 1.0 × 103 tons y−1 of Zn to the SYS sediments between 2011 and 2014, which accounted for 33% and 3% of total buried Hg and Zn, respectively. These findings indicate that environmental pollution control should also consider the oceanic contribution of some pollutants. The results of the present work help to elucidate the biogeochemical cycling of trace metals in marginal seas, and are helpful for managing environmental pollution in marine environments.

On the possibility to improve mixed uranium-plutonium fuel in fast reactors

Accumulation of plutonium extracted from the spent nuclear fuel (SNF) of light water reactors is one of the central problems in nuclear power. To reduce out-of-the-reactor Pu inventory, leading nuclear power countries (France, Japan) use plutonium in light water power reactors in the form of MOX fuel, with half of Pu fissioning in this fuel. The rest of Pu cannot be reused easily and efficiently in light water reactors because of the high content of even isotopes. Plutonium for which there are no potential consumers is accumulated. Unlike thermal reactors, fast reactors take plutonium of any isotopic composition. That makes it possible to improve plutonium isotopic composition and to reduce the fraction of even isotopes to the level that allows reuse of such plutonium in thermal reactors. The idea of changing the isotopic composition of Pu in fast reactors is well-known. The originality of the research lies in applying this idea to combine the fuel cycles of fast and thermal reactors. Pu isotopic composition can be improved by combining certain operational activities in order to supply fuel to thermal and fast reactors. Scientific and technological justification of the possibility will let Russian BN technologies and French MOX fuel technologies work in synergy with thermal reactors.

A highly‑enriched 244Pu reference material for nuclear safeguards and nuclear forensics measurements.

A highly-enriched 244Pu isotope dilution reference material has been prepared and characterized for metrologically traceable measurements of very small quantities of plutonium. The amount of plutonium in samples associated with nuclear safeguards and nuclear forensic measurements can be significantly less than 1 ng. Accordingly, the ability to quantify the amount and isotopic composition of plutonium from a single mass-spectrometric analysis is particularly desirable. The highly-enriched 244Pu reference material, described here, will minimize the magnitude of spike corrections necessary to obtain accurate information on plutonium isotopic composition from isotope dilution measurements.

Discrimination of weapons-grade plutonium from thermal reactors in nuclear forensics

A simulated fingerprinting approach is demonstrated, aiming the discrimination of weapons-grade plutonium (WG-Pu) from different reactors. Light and heavy water and graphite moderated thermal reactors are considered, charged with fresh fuel assemblies of different enrichments. The plutonium isotopic composition is considered as the required characteristic fingerprints in the approach. Hence, plutonium isotopic ratios are used in 3D isotopic correlations for the discrimination purposes. The WG-Pu from the different reactors is well resolved between them. Within the light water moderated cases, the WG-Pu is resolved in groups according to their fresh fuel enrichment and the reactor and fuel assembly type. The fingerprinting approach is sensitive to a ‘Pu age’ beyond 0.5 years.

Determination of 242Pu abundance in plutonium samples using self-fluorescent X-ray emission

Non-destructive gamma spectrometric measurements of the isotopic composition of plutonium are routinely carried out in the framework of nuclear safeguards inspections. High-resolution spectra of Pu obtained from planar Ge detectors are de-convoluted by means of special codes such as MGA or FRAM and these results are then applied when determining the total plutonium mass using neutron coincidence measurement techniques. The radiation emission of 242Pu is so weak that its abundance cannot be measured directly; hence isotopic correlations are utilized to estimate it in low and medium burn-up Pu material. The present work investigates the feasibility of determining the total Pu/Am ratio, exploiting X-ray fluorescence lines present in the plutonium spectrum. Combining the total Pu/Am ratio with individual Pu isotopic ratios to 241Am then allows an estimate of the 242Pu to be made. To this end, reference samples of plutonium, with an 241Am content comprised between 5 and 7 %, were measured using a planar germanium detector. The spectra were processed using MGA followed by quantification of the X-ray fluorescence peak of Americium, and then further calculations, including MCNP simulations. The conclusion of the present work is that, when the americium content of the sample is high enough to produce an Am Kα1 line which can be exploited, it is possible to indirectly determine the 242Pu abundance by analysing the X-ray fluorescent lines present in the plutonium gamma spectrum, thus raising the prospect to apply this method to cases for which the algorithms based on empirical correlations cannot be applied.

Plutonium assay and isotopic composition measurements in nuclear safeguards samples by inductively coupled plasma mass spectrometry

MC-ICP-MS methods for accurate and precise determination of Pu amount and isotopic composition at picogram to femtogram levels.

Identification of spent nuclear fuel with multivariate analysis based on database

In recent years, as the international situation about nuclear arms control changes, the multilateral international arms control cooperation including non-proliferation, nuclear terrorism and nuclear security has drawn more and more attention. As a key technology, nuclear forensics plays a significant role in the deterrence, prevention and response to illegitimate nuclear activities, for which it needs studying in depth. At present, there are plenty of researches into nuclear forensics, mostly focusing on the characterization of materials and the interpretation of data. As one of the most important aspects of nuclear forensics research, the interpretation is faced with a variety of different objects, including uranium ore, yellow cake, nuclear fuel, spent nuclear fuel and so on, among which spent nuclear fuel has attracted more and more attention due to its potential threats. In this paper, we primarily focus on the development of multivariate statistical analysis with an aim to interpret the comparative signatures of spent nuclear fuel. A database is established with uranium and plutonium isotopic compositions of spent nuclear fuel samples through simulation. These samples are of different reactor types, initial fuel enrichments and burn-ups. Subsequently, multivariate analysis, including factor analysis, discriminant analysis and regression analysis, are used to the database to validate the feasibility of the identification work. First of all, dimension reduction and visualization work is carried out to determine the possibility for classification by factor analysis. Afterwards, some known samples are assumed to be unknown to further study the possible capabilities of quantitative attribution by conducting factor analysis, including the determination of initial fuel enrichment and burn-up. To eliminate the errors in the identification work and to achieve better outcomes, the discriminant analysis and regression analysis are used to the database to assist with the identification of the reactor type, initial fuel enrichment and burn-up. As revealed by the study, factor analysis is more suitable for the dimension reduction and visualization work, disciminant analysis is more suitable for the identification of reactor type, and regression analysis is more suitable for the identification of initial fuel enrichment and burn-up. Upon the comparison drawn of the three different multivariate analysis methods, a framework for identification process is established to provide a theoretical basis and feasible scheme for the possible identification work of spent nuclear fuel with database, and facilitate the related nuclear forensics work.

Determination of the uranium enrichment without calibration standards using a 500 mm3 CdZnTe room temperature detector with a hybrid methodology based on peak ratios method and Monte Carlo counting efficiency simulations

Room temperature semiconductor detectors of CdZnTe type have been proposed as possible alternatives to spectrometers based on HPGe and NaI detectors. Attractive spectroscopic performance and portability, as well as continuous improvements in their design and availability in different sizes make them especially competitive in various radiation measurement applications where absence of cooling, small size and significantly better resolution compared to other room temperature detectors are an asset. Among these applications are nuclear safeguards that include uranium and plutonium isotopic composition determination tasks. This paper focuses on the investigation of possibilities and limits of a net peak area based methodology used for the determination of the uranium enrichment without use of calibration standards. Tests are conducted on a room temperature medium resolution spectrometer based on a 500 mm3 CdZnTe sensor of a quasi-hemispherical design using which spectra of different statistical quality are obtained from certified uranium standards. Gamma-ray peaks in the 143–1001 keV energy range are used as uranium gamma-ray signatures. Performance assessment of an intrinsic calibration of the counting efficiency curve is conducted. A hybrid methodology based on a combination of the peak ratios method with Monte Carlo simulations of the counting efficiency curve is proposed for poor quality spectra. Implemented algorithms and analysis routines are described in detail and presented.

Gamma-Spectrometric Control of Fissile Mass and Distribution Uniformity in Fuel Pins with Uranium-Plutonium Nitride Fuel

A non-destructive gamma-spectrometric method of determining the characteristics of fissile materials in fuel pins with uranium-plutonium fuel is presented. The measurements were performed with the aid of a scanning apparatus with an ultrapure germanium detector. The uniformity of the mass and plutonium distributions was monitored according to the counting rate of the pulses in the 413.7 keV peak of 239Pu. The FRAM code was used to determine the isotopic composition of plutonium in a fuel pin. Multiple repeated measurements were performed and the repeatability and reproducibility of the determinations of the characteristics of the fissile materials in a fuel element were determined.

Identification of unknown spent nuclear fuel with factor analysis for nuclear forensic purpose

The proposed study focuses on the development of multivariate statistical analysis to interpret the comparative signatures of spent nuclear fuel. A database was built with Uranium and Plutonium isotopic compositions of spent nuclear fuel samples through simulation. These samples are in different reactor types, initial fuel enrichments and burn-ups. As a preliminary study, factor analysis is applied to the database to certify the feasibility of the possible origin-determination work. At first, dimension reduction and visualization work is realized to decide the possibility for classification. Then, some known samples are assumed to be unknown to further study the possible capabilities for quantitative attribution, containing the determination of initial fuel enrichment and burn-up.

Supported liquid membrane based loading technique for thermal ionization mass spectrometry: an application to plutonium isotopic composition and concentration determination

Supported liquid membrane (SLM) was prepared by immobilizing the extractant tricaprylmethyl ammonium chloride (Aliquat-336) into the microporous poly(propylene) membrane. The formed SLM with optimize composition was found be selective toward Pu4+ ions over UO22+ and Am3+ ions in a wide range of high acid concentrations (3–7 mol L−1 HNO3). The composition of SLM was optimized by using varying proportions of Aliquat-336 with solvents/diluents such as dioctyl phthalate (DOP), 2-nitrophenyl octyl ether and tris(2-ethylhexyl) phosphate. Among these diluents, DOP was found to be stable and causing uniform distribution of the extractant on the membrane. The surface morphology was characterized by atomic force microscopy. Stability and change in physical structure of SLM was characterized by capillary flow porometry. The Pu pre-concentrated in SLM was used to find out isotopic composition and concentration of the plutonium with help of thermal ionization mass spectrometry. The SLM was found to be stable and amenable to routine analyses.