Short-lived Fission Products Research Articles

Toward short-lived and energy-dependent fission product yields from neutron-induced fission

Fission product yields (FPYs) are an important source of information that are used for basic and applied physics. They are essential observables to address questions relevant to nucleosynthesis in the cosmos that created the elements from iron to uranium, for example, in energy generating processes from fission recycling in binary neutron star mergers; resolving the reactor neutrino anomaly; decay heat release in nuclear reactors; and many national security applications. While new applications will require accurate energy-dependent FPY data over a broad set of incident neutron energies, the current evaluated FPY data files contain only three energy points: thermal, fast, and 14-MeV incident energies.Recent measurements using mono-energetic and pulsed neutron beams at the Triangle Universities Nuclear Laboratory (TUNL) tandem accelerator and employing a dual fission ionization chambers setup have produced self-consistent, high-precision data critical for testing fission models for the neutron-induced fission of the major actinide nuclei. This paper will present new campaign just beginning utilizing a RApid Belt-driven Irradiated Target Transfer System (RABITTS) to measure shorter-lived fission products and the time dependence of fission yields, expanding the measurements from cumulative towards independent fission yields.

Fission Product Yield Measurements from Neutron-Induced Fission of 235,238 U and 239Pu

Fission product yields (FPY) are one of the most fundamental quantities that can be measured for a fissioning nucleus and are important for basic and applied nuclear physics. Recent measurements using mono-energetic and pulsed neutron beams generated using Triangle Universities Nuclear Laboratory’s tandem accelerator and employing a dual fission chamber setup have produced self-consistent, high-precision data critical for testing fission models for the neutron-induced fission of 235,238U and 239Pu between neutron energies of 0.5 to 15.0 MeV. These data have elucidated a low-energy dependence of FPY for several fission products using irradiations of varying lengths and neutron energies. This paper will discuss new measurements just beginning utilizing a RApid Belt-driven Irradiated Target Transfer System (RABITTS) to measure shorterlived fission products and the time dependence of fission yields, expanding the measurements from cumulative towards independent fission yields. The uniqueness of these FPY data and the impact on the development of fission theory will be discussed.

Characterization of californium sources by gamma spectrometry: relevance for nuclear forensics

High-resolution gamma spectrometry was used to study characteristics of californium neutron sources—age, isotopic composition and impurity radionuclides, as well as the effect of the isotopic composition on uncertainties involved in age determination. These characteristics are relevant during criminal investigations related to californium sources found outside of regulatory control. Age determination relies on measuring the activity ratios of long-lived and short-lived fission products of 250Cf and 252Cf. It is based on a known method but also introduces new pairs of isotopes that can be used for this purpose. The isotopic composition was estimated from the direct measurement of the activities of 249Cf and 251Cf isotopes. The methods were tested using 5 californium reference sources at JRC-Karlsruhe. Additional raw data (gamma spectra) were provided by the German Radiation Protection Authority, and by the University of Novi Sad, Serbia. It has been shown that gamma spectrometry can be successfully used to characterize californium sources for nuclear forensic purposes.

Model design of a compact delayed gamma-ray moderator system using 252Cf for safeguards verification measurements

Delayed gamma-ray spectroscopy (DGS) is an active-interrogation nondestructive assay technique that can be used to determine the composition of a nuclear material sample by comparing the ratios of fission product gamma-ray peak intensities. However, high-radioactivity nuclear material (HRNM) contains long-lived fission products that can overwhelm a detector and shielding must be used to reduce the count rate, while minimally affecting those gamma rays from short-lived fission products with energy above 3-MeV. To compensate for the signal loss through the shielding, low-energy neutrons are required to induce more fission events from the high thermal cross-section of fissile nuclides. To improve practical safeguards DGS capabilities, we are developing a compact interrogation system to moderate ∼ 2-MeV neutrons that are easier to moderate than 14-MeV neutrons from standard deuterium-tritium generators. This work describes the optimization of an ideal moderator system for a 252Cf neutron point source that results in a neutron fluence of 25.9×10−4 cm−2nsource−1 passing through the sample space with ≳ 70% of those below 1-eV. Modifications for practical fabrication resulted in ⩽20% reductions of the flux compared to the optimized ideal design. Finally, evaluations made of HRNM DGS signals and backgrounds conclude that a 252Cf source intensity of 8.9×107 neutrons per second is required for a single-pass interrogation within this ideal moderator system. However, this can be as low as 3.6×106 neutrons per second using smaller samples that require less shielding.

Thermal Stability of Glass with Simulators of Chloride Highly Radioactive Wastes

High temperature behavior of sodium–aluminum fluorophosphate glass—a potential matrix for immobilisation of waste salt electrolyte from pyrochemical reprocessing of irradiated nuclear fuel (INF) has been studied. The glass crystallizes between 430 and 640°C with formation of phosphate phases, which dissolve above 640°C leading to homogenization of the glass. Similar transformations of the glass matrix due to heating from the decay of short-lived fission products may have a negative effect on properties of vitrified radioactive chloride wastes during storage in a repository.

Radiochemical separations and experimental measurements of short-lived fission products from 14 MeV irradiation of depleted uranium

This work focused on the experimental analysis of depleted uranium irradiated with 14 MeV neutrons to obtain fission yields for 20 fission products with half-lives ranging from 16.8 h to 30.1 years. Several analytes were identified where improvements to the available nuclear data would be beneficial. The results obtained from this experiment could be used to make improvements to cumulative fission yield data once confirmatory measurements are made. The required separations will be described along with a comparison to the reported Evaluated Nuclear Data File fission yields and other experimental measurements.

Neutron activation of NIST surrogate post-detonation urban debris (SPUD) candidate SRMs

Despite their importance, there is a dearth of post-detonation nuclear forensics standard reference materials (SRMs) suitable for analysis traceable back to a national standard. Accordingly, the nuclear forensics community has requested SRMs be produced that mimic post-detonation fallout debris that include actinides, urban materials, fission products, and activation products. The National Institute of Standards and Technology (NIST), in partnership with the National Physical Laboratory and with support from the Federal Bureau of Investigation, have developed two surrogate post-detonation urban debris (SPUD) candidate SRMs to mimic the “rubble” of a city after an improvised nuclear device detonation. NIST SPUD samples were irradiated at The University of Texas TRIGA reactor, then analyzed via gamma-ray spectroscopy for short-lived, medium-lived, and long-lived fission and activation products.

Depletion measurements using a new gamma transparency method of irradiated MTR-type fuel assemblies

An innovative method for the determination of the depletion in research reactor fuel assemblies (FAs) based on gamma transparency was developed, tested, and presented. The basic concept is the assessment of the uranium content from the attenuation of gamma rays passing through the FA, which is placed between a gamma-emitting source and a spectroscopic detector. With increasing depletion, more uranium transforms into fission products, which causes a significant decrease in the gamma-rays attenuation at energies higher than ∼116 keV, thus increasing the count rate at the detector. The attenuation from the fuel assembly was compared to that of a fresh (highest known uranium content, most opaque to gamma-rays) and dummy (Aluminum-only; most transparent) fuel assemblies, and subsequently, a logarithmic interpolation to determine the current uranium content was performed. In this study, a low-energy gamma ray of 155 keV (emitted from 188Re) was used, which exhibited significant contrast between the attenuation through the uranium versus its fission products, while obtaining sufficiently high count-rate at the detector at relatively short count durations. The 188Re source was generated prior to the experimental campaign through the 187Re(n,γ) reaction by irradiation of natural rhenium in IRR1 (Israel Research Reactor 1, 5MW open-pool HEU MTR-type fuel, Soreq Nuclear Research Center, Israel). Five FAs of different irradiation histories previously used in the IRR1 reactor were studied. The determined depletions were found to be in the range of 47%–59%, with absolute uncertainties of 2%–3%. The axial distribution of the depletion along the FAs was determined with high resolution (20 axial points per FA) using the low-energy photons and found to be in good agreement compared to the axial distributions obtained from fission products such as 137Cs. Experimental results of the present study were also compared to results of a different experiment that used the traditional gamma ray spectroscopy technique (i.e., using 137Cs activity) for one of the studied fuel assemblies (FS-5), giving good agreement in both the axial depletion distribution, as well as the whole assembly depletion. Thus, the method described in this paper demonstrates improved accuracy compared to other non-destructive spectroscopy-based methods. Furthermore, we managed to obtain spectroscopic signatures of relatively short-lived fission products such as 140Ba (T1∕2=12.7 days) from an FA that was irradiated 65 days before the measurement.

One candidate method based on short-lived indicator for burnup analysis

One novel gamma spectrometry method based on short-lived fission products was developed that can infer the level of burnup without complex computations or knowing irradiation history. The relationship between the burnup information and equilibrium concentration of short-lived indicators was derived. In addition, new candidate indicators for gamma spectrum analysis, such as 88Kr, 91Sr, 92Sr, 135I and 142La were screened out. The results of three cases simulated with ORIGEN and MCNP were in good agreement with the theoretical values.

Cu metal based radiotellurium separation from simulated and HEU process streams

Radioiodine interferes significantly in short-lived and low yield fission product detection by radiometric means. A simple method has been developed for removing radiotellurium and its radioiodine daughters from process HEU analyte streams. The method uses metallic copper relying on the electrochemical behavior between Cu and Te to deposit Te on the surface of Cu packed chromatography columns. Method development on stable metal samples exhibited > 93% Te removal from high HCl conditions (> 5 M). The method was applied to processed HEU streams providing > 90% Te removal (examining 129mTe and 132Te), and providing > 99.9% radioiodine removal (examining 131I and 132I).

An optimization methodology for heterogeneous minor actinides transmutation

In the case of a closed fuel cycle, minor actinides transmutation can lead to a strong reduction in spent fuel radiotoxicity and decay heat. In the heterogeneous approach, minor actinides are loaded in dedicated targets located at the core periphery so that long-lived minor actinides undergo fission and are turned in shorter-lived fission products. However, such targets require a specific design process due to high helium production in the fuel, high flux gradient at the core periphery and low power production. Additionally, the targets are generally manufactured with a high content in minor actinides in order to compensate for the low flux level at the core periphery. This leads to negative impacts on the fuel cycle in terms of neutron source and decay heat of the irradiated targets, which penalize their handling and reprocessing. In this paper, a simplified methodology for the design of targets is coupled with a method for the optimization of transmutation which takes into account both transmutation performances and fuel cycle impacts. The uncertainties and performances of this methodology are evaluated and shown to be sufficient to carry out scoping studies. An illustration is then made by considering the use of moderating material in the targets, which has a positive impact on the minor actinides consumption but a negative impact both on fuel cycle constraints (higher decay heat and neutron) and on assembly design (higher helium production and lower fuel volume fraction). It is shown that the use of moderating material is an optimal solution of the transmutation problem with regards to consumption and fuel cycle impacts, even when taking geometrical design considerations into account.

On the influence of the americium isotopic vector on the cooling time of minor actinides bearing blankets in fast reactors

In the heterogeneous minor actinides transmutation approach, the nuclei to be transmuted are loaded in dedicated targets often located at the core periphery, so that long-lived heavy nuclides are turned into shorter-lived fission products by fission. To compensate for low flux level at the core periphery, the minor actinides content in the targets is set relatively high (around 20 at.%), which has a negative impact on the reprocessing of the targets due to their important decay heat level. After a complete analysis of the main contributors to the heat load of the irradiated targets, it is shown here that the choice of the reprocessing order of the various feeds of americium from the fuel cycle depends on the actual limit for fuel reprocessing. If reprocessing of hot targets is possible, it is more interesting to reprocess first the americium feed with a high243Am content in order to limit the total cooling time of the targets, while if reprocessing of targets is limited by their decay heat, it is more interesting to wait for an increase in the241Am content before loading the americium in the core. An optimization of the reprocessing order appears to lead to a decrease of the total cooling time by 15 years compared to a situation where all the americium feeds are mixed together when two feeds from SFR are considered with a high reprocessing limit.

CAD modeling study on FLUKA and OpenMC for accelerator driven system simulation

Accelerator driven system (ADS) is an advanced nuclear system with high inherent safety feature, which can transmute the nuclear waste including minor actinides (MA) and long-lived fission products (LLFPs) into short-lived fission products. In order to accomplish the neutronic simulation in ADS, a conventional way can be divided into two steps, namely the physical process of proton beam bombarding spallation target (FLUKA) and the neutron transport process in subcritical reactor coupling with a spallation neutron source (OpenMC). However, there are some difficulties in the modeling process for ADS, such as the inconsistent modeling process between FLUKA and OpenMC codes, and the error-prone process of constructing complex geometries. Therefore, it is imperative to build a code system that can convert the CAD modeling files into the input cards of Monte Carlo codes. In this context, a code system named CAD-PSFO (FreeCAD based parsing script for FLUKA and OpenMC) has been developed to solve the modeling conversion problem in ADS, in which basic geometry and boolean logic classes have been established with a mapping relationship to FLUKA and OpenMC codes, and ray-casting technology based iteration method has been proposed to solve the problem of higher-order surfaces in complex geometries. Finally, CAD-PSFO has been verified and validated by comparing with a reference model, two benchmarks, and two simulation tests.

Optimization of minor actinides bearing blankets with industrial fuel cycle constraints

Heterogeneous minor actinides transmutation is the process of loading neptunium, americium and curium in dedicated targets, generally located at the core periphery, in order to transmute these long-lived heavy nuclides into shorter-lived fission products. This has a positive effect on long-term radiotoxicity of the spent fuel and on the short to medium term decay heat of the final waste packages. However, given their location at the core periphery in a low flux zone, these targets are generally loaded with a high content in minor actinides in order to maintain acceptable transmutation performances. A significant part of the heavy nuclides produced by capture during irradiation being energetic alpha-emitters or strong neutron emitters, this leads to a very high activity in the irradiated targets which has negative impacts on the fuel cycle. In this paper, an optimization methodology of heterogeneous minor actinides transmutation is applied to the fuel cycle case in order to assess the constraints limiting transmutation. It appears that transportation of the fresh assemblies is the first limiting criterion due to the high decay heat of 241Am. Sodium washing at the end of irradiation is the second limiting criterion. It is shown that the geometrical design on the target assembly is not a concern considering the current limitations on the fuel cycle. It is also shown that using current industrial limitations and plutonium feed, it is not possible to achieve equilibrium between production of americium in the core and consumption in the blankets. It appears then that research should focus on the removal of fuel reprocessing limitations rather than standard assembly optimization.

Modelling the external radiation exposure from the Chernobyl fallout using data from the Swedish municipality measurement system

In connection with the Chernobyl fallout and the subsequent deposition of radionuclides in Sweden, Swedish municipalities launched a measurement program to monitor the external radiation exposure. This program encompasses measurements of the ambient dose equivalent rate 1 m above ground at selected locations, and repeats those measurements at the same locations at 7-month intervals. Measurement data compiled from the seven locations with the highest deposition were combined with data from aerial surveys since May 1986 of ground deposition of 137Cs, high-resolution gamma spectrometry performed at four locations in May 1986, and measurements from fixed continuous air gamma rate monitoring stations from 28 April to 15 May 1986. Based on these datasets, a model of the time pattern of the external dose rate in terms of ambient dose equivalent rate from the Chernobyl fallout was developed. The decrease in the ambient dose equivalent rate could, on average, be described by a four-component exponential decay function with effective half-times of 6.8 ± 0.3 d, 104 ± 26 d, 1.0 ± 0.02 y and 5.5 ± 0.09 y, respectively. The predominant contributions to the external dose rate in the first month were from short-lived fission products superseded by 134Cs and then 137Cs. Integrated over 70 y and using extrapolation of the curve fits, our model predicts that 137Cs contributes about 60% and 134Cs contributes about 30% of the external effective dose at these seven locations. The projected time-integrated 70 y external effective dose to an unshielded person from all nuclides per unit total activity deposition of 137Cs is estimated to be 0.29 ± 0.0.08 mSv/(kBq m−2). These results are in agreement with those found in Chernobyl contaminated Russian forest areas, and emphasize the usefulness of maintaining a long-term and regular measurement program in contaminated areas.

Fission Product Yields from 232Th, 238U, and 235U Using 14 MeV Neutrons

Neutron-induced fission yield studies using deuterium-tritium fusion-produced 14 MeV neutrons have not yet directly measured fission yields from fission products with half-lives on the order of seconds (far from the line of nuclear stability). Fundamental data of this nature are important for improving and validating the current models of the nuclear fission process. Cyclic neutron activation analysis (CNAA) was performed on three actinide targets–thorium-oxide, depleted uranium metal, and highly enriched uranium metal–at the University of Michigan's Neutron Science Laboratory (UM-NSL) using a pneumatic system and Thermo-Scientific D711 accelerator-based fusion neutron generator. This was done to measure the fission yields of short-lived fission products and to examine the differences between the delayed fission product signatures of the three actinides. The measured data were compared against previously published results for 89Kr, −90, and −92 and 138Xe, −139, and −140. The average percent deviation of the measured values from the Evaluated Nuclear Data Files VII.1 (ENDF/B-VII.1) for thorium, depleted-uranium, and highly-enriched uranium were −10.2%, 4.5%, and −12.9%, respectively. In addition to the measurements of the six known fission products, 23 new fission yield measurements from 84As to 146La are presented.

Measurements of extinct fission products in nuclear bomb debris: Determination of the yield of the Trinity nuclear test 70 y later

This paper describes an approach to measuring extinct fission products that would allow for the characterization of a nuclear test at any time. The isotopic composition of molybdenum in five samples of glassy debris from the 1945 Trinity nuclear test has been measured. Nonnatural molybdenum isotopic compositions were observed, reflecting an input from the decay of the short-lived fission products (95)Zr and (97)Zr. By measuring both the perturbation of the (95)Mo/(96)Mo and (97)Mo/(96)Mo isotopic ratios and the total amount of molybdenum in the Trinity nuclear debris samples, it is possible to calculate the original concentrations of the (95)Zr and (97)Zr isotopes formed in the nuclear detonation. Together with a determination of the amount of plutonium in the debris, these measurements of extinct fission products allow for new estimates of the efficiency and yield of the historic Trinity test.

A generalized method for characterization of 235U and 239Pu content using short-lived fission product gamma spectroscopy

Rapid nondestructive assay methods for trace fissile material analysis are needed in both nuclear forensics and safeguards communities. To address these needs, research at the Oak Ridge National Laboratory High Flux Isotope Reactor Neutron Activation Analysis facility has developed a generalized nondestructive assay method to characterize materials containing fissile isotopes. This method relies on gamma-ray emissions from short-lived fission products and makes use of differences in fission product yields to identify fissile compositions of trace material samples. Although prior work has explored the use of short-lived fission product gamma-ray measurements, the proposed method is the first to provide a complete characterization of isotopic identification, mass ratios, and absolute mass determination. Successful single fissile isotope mass recoveries of less than 6% recovery bias have been conducted on standards of 235U and 239Pu as low as 12ng in less than 10 minutes. Additionally, mixtures of fissile isotope standards containing 235U and 239Pu have been characterized as low as 198ng of fissile mass with less than 7% recovery bias. The generalizability of this method is illustrated by evaluating different fissile isotopes, mixtures of fissile isotopes, and two different irradiation positions in the reactor. It is anticipated that this method will be expanded to characterize additional fissile nuclides, utilize various irradiation facilities, and account for increasingly complex sample matrices.

Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

This study presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t 1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (>5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

Reducing uncertainties for short lived cumulative fission product yields

Uncertainties associated with short lived (half-lives less than 1 day) fission product yields listed in databases such as the National Nuclear Data Center’s ENDF/B-VII are large enough for certain isotopes to provide an opportunity for new precision measurements to offer significant uncertainty reductions. A series of experiments has begun where small samples of 235U are irradiated with a pulsed, fission neutron spectrum at the Nevada National Security Site and placed between two broad-energy germanium detectors. The amount of various isotopes present immediately following the irradiation can be determined given the total counts and the calibrated properties of the detector system. The uncertainty on the fission yields for multiple isotopes has been reduced by nearly an order of magnitude.