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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Angular Correlations of Gamma Rays from Fission Product Solution

Fission products from photon-induced fission are routinely used at Idaho National Laboratory as part of nuclear forensics activities. Typically, the focus is on fission products with relatively large fission yields (Ba-140, Mo-99, etc.). The nuclear data associated with the well-known isotopes have small uncertainties, and many have well-known angular correlations between emitted gamma-rays, if applicable. However, for isotopes with smaller fission yields or less intense gamma-ray emission intensities, the nuclear decay schemes and gamma-ray emission intensities may be less well-known. The Germanium Rotational Measurements for Angular Correlation (GeRMAC) system was designed to use two high-purity germanium (HPGe) detectors in a coincidence counting set-up but allows one of the HPGe detectors to rotate at a constant distance from a radioactive source, changing the relative angle between the two HPGe detectors. Angular correlations can then be determined from the relative photopeak areas between two gamma-rays emitted as part of the same nuclear decay. By recording the HPGe spectra in list mode, the angular correlations of gamma-rays emitted from complex sources, consisting of multiple isotopes, can be measured simultaneously and the individual isotope emissions can be isolated. Using the GeRMAC system, a multi-isotope fission product source, produced with photon-induced fission of U-238, was measured using a single rotational arc (180 degrees) to obtain the angular correlation between gamma-rays emitted by the large number of fission products contained in the solution with an angular accuracy of 1 degree. In particular, La-140 (38 gamma rays) and Nd-147 (26 gamma rays) are of interest to the nuclear forensics community.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2020

Measurement of distribution coefficients on TODGA resin and development of a group separation for mixed activation and fission product samples

Batch equilibrium experiments were performed to determine the distribution coefficients for Sc, Ir, and Ta on N,N,N’,N’tetraoctyl-1,5-diglycolamide (TODGA) resin in various acid matrices. Distribution coefficients were determined in hydrochloric acid (Sc, Ir), nitric acid (Sc, Ta), and a mixture of nitric acid and 0.2M hydrofluoric acid (Sc, Ir, Ta). The distribution coefficients determined through these experiments were then utilized to create a group separation method for activation products, fission products, and environmental constituents.

Louie, Christina A.↗
NASA NTRS1967

Computer program FPIP-REV calculates fission product inventory for U-235 fission

Computer program calculates fission product inventories and source strengths associated with the operation of U-235 fueled nuclear power reactor. It utilizes a fission-product nuclide library of 254 nuclides, and calculates the time dependent behavior of the fission product nuclides formed by fissioning of U-235.

Brown, W. S.↗
DOE OSTI2020

Zinc-in-titania waste form for immobilizing lanthanide fission products from electrochemical reprocessing

This paper provides microstructural, microchemical, and bulk X-ray diffraction information for a zinc-in-titania waste form for immobilizing rare-earth oxide (REOx) fission products recovered through precipitation from electrochemical salt wastes. This waste form can be used to immobilize REOx at loadings as high as 40 mass% in a chemically durable matrix consisting of glass-bonded minerals including Zn2TiO4 as well as RE-containing phases of REOx, RE2Ti2O7, and REPO4 (monazite). Chemical durability data is also provided, and comparisons are drawn between this waste form and one made using a similar process with half the REOx loading.

rare earth, rare earth waste form, electrochemical↗
DOE OSTI2024

Energy dependence of fission product yields in the second-chance fission region

An extensive dataset of cumulative fission product yields has been generated under a joint collaboration between Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and the Triangle Universities Nuclear Laboratory (TUNL). The energy dependence of the cumulative yield for a select number of high-yield fission products has been measured using quasimonoenergetic neutrons with energies between 5.5 and 11.0 MeV. This is in addition to previously published data covering 0.5–4.5 and 14.8 MeV. The absolute number of fissions was determined during the irradiation period using dual-fission ionization chambers, and the fission products were measured postactivation by whole target γ-ray spectroscopy. Here this paper presents the absolute cumulative fission product yields as a function of incident neutron energy from the neutron-induced fission of 235 U, 238 U, and 239 Pu isotopes at four incident energies in the second-chance fission region and compares them with existing literature values. Corrections relevant to this collaboration's previously published data are also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Characterization of a platform for the gas transport, collection, and identification of fission products in the high-intensity laser environment

Recent progress in the production of laser accelerated high flux proton beams opens new possibilities for the study of fission in unique environments. To this end, we are currently pursuing the development of a platform for the swift gas transport, collection, and identification of fission products. Fission is induced in targets fixed inside a sealed chamber through which a carrier gas flows, transporting fission products to a carbon filter for collection and online spectroscopy. This has been recently demonstrated using target normal sheath accelerated protons at the PHELIX laser facility. There were large discrepancies between measured rates and those expected based on established fission yields and measured beam parameters. These discrepancies prompted a large number of tests at the Idaho Accelerator Center (IAC), where fission in uranium is induced by bremsstrahlung photons generated by 21 MeV electrons from a linear electron accelerator. Here, the results are compared to a series of models that account for the slowing of energetic fission fragments in the carrier gas and the fluid dynamics of the gas flow that transports fission products to the collection filter. Characterization of the apparatus reveals a few mechanisms that together account for a portion of the previously observed discrepancies at the PHELIX laser facility. However, additional research is necessary before high-accuracy experiments can be performed with the apparatus. Work performed under the auspices of the U.S. Department of Energy by LLNL under contract DE-AC52-07NA27344.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Separation of Lanthanide Isotopes from Mixed Fission Product Samples

The measurement of radioactive fission products from nuclear events has important implications for nuclear data production, environmental monitoring, and nuclear forensics. In a previous paper, the authors reported the optimization of an intra-group lanthanide separation using LN extraction resin from Eichrom Technologies®, Inc. and a nitric acid gradient. In this work, the method was demonstrated for the separation and quantification of multiple short-lived fission product lanthanide isotopes from a fission product sample produced from the thermal irradiation of highly enriched uranium. The separations were performed in parallel in quadruplicate with reproducible results and high decontamination factors for 153 Sm, 156 Eu, and 161 Tb. Based on the results obtained here, the fission yields for 144 Ce, 153 Sm, 156 Eu, and 161 Tb are consistent with published fission yields. This work demonstrates the effectiveness of the separations for the intended application of short-lived lanthanide fission product analysis requiring high decontamination factors.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Development and application of sorption mass transfer models for fission product transport in TRISO fuel systems using BISON

Fission product mass transfer within and between TRISO particle and compact layers is an important phenomenon. It directly impacts fission product release predictions, which are used as source terms for safety and licensing calculations. Modeling fission product transport within layers and across bonded interfaces is relatively straightforward under the assumptions of isotropic Fickian diffusion, concentration continuity, and flux continuity. Modeling fission product transport across gaps and debonded layers is more difficult. Gaps are known to form between the buffer and inner PyC (IPyC), and debonding may occur between the IPyC and silicon carbide (SiC). A new mass transfer model was developed to provide a more accurate and robust fission product release calculation by accounting for interlayer sorptivity. One importance of the new model is to account for temperature and material changes across the gap based on sorption isotherm. This work presents the development of the sorption behavioral model and a fission product trapping model, and analyses the subsequent fission product diffusion behavior, specifically cesium, through TRISO particles and compact materials using the BISON fuel performance code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Fission Product Yield Modeling and Evaluation

Although independent and cumulative fission product yields have been a part of evaluated libraries for decades, there have been few updates over the years. The fission product yield sub-library in the ENDF/B-VIII.0 library is still largely based on the evaluation of England and Rider from the mid-90’s, with only more recent updates to the energy dependence of 239 Pu below 2 MeV and fixes to isomeric states and missing fission products. Over the past several years, there have been a wealth of new measurements of independent and cumulative fission product yields, particularly those with short half-lives, and there have been significant improvements in the modeling of prompt and delayed fission observables. Here, we describe recent progress in the improvement of fission product yield calculations, using the BeoH code and the underlying Hauser Fesh-bach Fission Fragment Decay (HF 3 D) model, developed at Los Alamos National Laboratory. We will describe our recent calculations for consistent prompt and delayed fission observables for major and minor actinides, including new work investigating isomeric ratios. We will detail the ongoing evaluation process for energy-dependent fission product yields from thermal up to 20 MeV incident neutron energy and some validation work that has been performed for these new fission product yield calculations. Additionally, we will discuss future perspectives of this work, highlighting the need for additional data.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Micro- and Nano-techniques for the Study of Fission Product Precipitation in SiC Layer

The complexity of fission product distribution and composition within the SiC layer of tristructural isotropic (TRISO) coated particles, irradiation effects on the SiC structure as well as the variable nature of metallic fission product release, has led to the exploration of multiple micro- and nano-characterization techniques. The fine scale of the fission product precipitates necessitates unique learnings and application of electron microscopic techniques with irradiated fuel during the Advanced Gas Reactor (AGR)-1 experiment analysis. A summarized discussion on the specific advanced techniques with associated method development for TRISO coated particles is provided followed by the down selected techniques currently considered to provide the highest impact. Techniques considered are electron probe micro-analyzer (EPMA), scanning transmission electron microscopy (STEM), and Precession electron diffraction (PED). SiC grain boundary characteristics are evaluated due to the mobility of specific fission products through grain boundaries. PED was performed for understanding the role of grain boundary character on fission product transport. One set of comparative analyses between AGR-1 and AGR-2 particles leads to a finding that the AGR-1, high-Ag-retention particle, had statistically more coincidence site lattice (CSL)-related grain boundaries (but fewer low-angle grain boundaries), compared to the AGR-2 particle. This implies that CSL-related grain boundaries may have a direct influence on Ag retention. However, precipitates were found on a relatively small fraction of CSL-related grain boundaries in both particles. In another set of comparison between AGR-1 and AGR-2 particles that were subjected to safety testing, there is significant differences in average SiC grain boundary distributions, but the fission product distributions on these grain boundaries are very similar.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

AGR-5/6/7 Irradiation Experiment Fission Product Mass Balance

This report presents the fission product mass balance for the AGR-5/6/7 TRISO fuel irradiation experiment. The fission product inventories deposited on capsule components outside of the fuel (e.g., stainless-steel shells, graphite holders, Grafoil disks, and associated hardware) were quantified as part of the post-irradiation examination (PIE) to assess the performance of this fuel. Comparisons were made between these inventories and depletion calculations, non-destructive measurements of fuel fission product inventories in the intact fuel compacts, prior AGR experiments, and results from among each of the five distinct AGR-5/6/7 capsules. The data served as estimates of the condensable fission products released from the fuel during irradiation. Excluding Capsule 1 (which experienced accidental damage during irradiation) and Capsule 3 (which was tested at very high irradiation temperatures of >1300°C), the results indicate that the AGR-5/6/7 fuel performed comparably to fuel from earlier AGR experiments. The mass balance results were also used to estimate the number of particles with in-pile SiC failures. Subject to the assumptions made in these estimates, and excluding Capsules 1 and 3, the in-pile SiC failure rates for AGR-5/6/7 are comparable to those observed in AGR-2.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

AGR TRISO Fuel Fission Product Release Data Summary

Knowledge of fission product retention in and release from TRISO fuel under normal and off-normal conditions is needed for reactor safety analyses. This is important for coated-particle fuel used in high-temperature reactors relying on the functional containment strategy. Data on the release and retention of key fission products (e.g., Ag-110m, Cs-134, Eu-154, and Sr-90) in AGR UCO TRISO fuels have been summarized in this report, and empirical relationships with respect to time and temperature were developed. This included fission product accumulation in the OPyC and compact graphitic matrix during irradiation, release from compacts during irradiation, and release during post-irradiation safety testing at temperatures from 1600-1800°C. The frequencies of SiC failure and TRISO coating failure from irradiation and post-irradiation safety testing were also summarized as they have bearing on the quantities of Cs release from the fuel. In a forthcoming publication, a framework for combining and using these empirical relationships as part of a source term analysis will be presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Energy Dependent Fission Product Yields (2nd year Milestone and Deliverables)

Introduction to the project: One of the Laboratory missions is to provide consistent, high-precision fission product; yield data critical for testing fission models and maintaining the safety and security of the nation’s nuclear weapons stockpile. However, high-quality, energy-differential fission product yield data is missing for certain actinides and neutron energies important to constrain the new U.S. Nuclear Data Program evaluation effort; It has been shown that the reactor antineutrino anomaly may be at least partially caused by roughly 20 fission products. The fission product yield data is missing or incomplete for many of these isotopes, thus it is necessary to accurately determine these values to better constrain the anomaly; Aitor Bracho is measuring very short-lived (seconds to minutes) fission product yields of 235 U and 239 Pu using monoenergetic neutron beams at E n = 60 and 560 keV; Aitor Bracho is using a direct approach utilizing a state-of-the-art rabbit transfer system, superior HPGe detector, and digital acquisition systems for fission decay measurements. The goals of this project: lop experimental capabilities and data analysis techniques to carry out the gamma-ray spectra analysis necessary for fission product yield calculation; Provide high-precision and energy-dependent fission product data supporting fission theory, neutrino physics, and applied physics.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Energy dependence of chain fission product yields from neutron-induced fission of 235 U, 238 U, and 239 Pu

The lack of completeness and systematic studies of cumulative fission product yield (FPY) from the three major actinides of 235 U, 238 U, and 239 Pu, over a broad energy range, was the primary motivation for the LLNL-LANL-TUNL collaboration. Given the focus on resolving FPY behavior around fission spectrum neutron energies, our 2016 data provides FPY measurements in fine energy steps in the range of 0.5–5.5 MeV and an additional anchor point at 14.8 MeV. Recently, our collaboration filled the gap between second chance fission and 14.8 MeV, providing a comprehensive study at eleven incident neutron energies from 0.5 to 14.8 MeV. However, during the course of the last measurements, it was found that the reported masses of the actinide deposits in the fission ionization chambers, which were used to measure the number of fissions, were not consistent with the complementary tests. Therefore, some corrections to the old data are needed. The purpose of this paper is to summarize all of the high-yield FPY data obtained for 235 U, 238 U, and 239 Pu isotopes using quasi-monoenergetic neutron beams at 11 incident energies ranging from E n = 0.5 to 14.8 MeV.

Tonchev, A. P. [Lawrence Livermore National Labora↗
DOE OSTI2021

Modeling fission product diffusion in TRISO fuel particles with BISON

Diffusion of fission products in intact TRISO particles depends on particle geometry, fission product source rates, time, temperature, and temperature-dependent diffusion coefficients. Simulating this diffusion process requires models for source rates and diffusion coefficients, plus computation of the temperature field if not prescribed. In addition, simulation quality depends on discretization of the geometry, appropriate time stepping, and the accuracy of the solution method. In this paper, we explore the simulation of fission product diffusion in TRISO fuel particles using the finite element method via the fuel performance code Bison. Recent material model development has occurred in Bison for each material present in tri-structural isotropic (TRISO) fuel particles: the buffer, inner pyrolytic carbon, silicon carbide, and outer pyrolytic carbon layers, as well as the fuel kernel. Also, new mesh generation and fission product release fraction capabilities have been added. Diffusion capabilities are shown to converge to the correct solution via formal verification tests. A large number of code benchmarking problems are also given, with good results, showing that Bison’s computed release fractions closely match those of other software tools. Finally, a significant validation effort is detailed in which fission product release, measured as part of the AGR-1 capsule experiments, is compared to Bison outputs. Bison outputs compare very well to the experimental data and to PARFUME results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Measurement of Fission Product Concentration Profiles in AGR-3/4 TRISO Fuel Graphitic Matrix and Nuclear Graphites

The third Advanced Gas Reactor (AGR) irradiation experiment, AGR-3/4, was designed to investigate the migration of fission products in fuel compact graphitic matrix and reactor graphite components. Using destructive methods, radial fission product concentration profiles were measured for gamma-emitting fission products (e.g., Ag-110m, Cs-134, and Eu-154) and beta-emitting Sr-90 in irradiated graphitic and graphite components from six different AGR-3/4 irradiation capsules. These new measured concentration profiles can now be compared to non-destructive measurements and fission product transport simulations and will be used to derive new diffusivities and sorptivities to support refinement of fission product transport models and high-temperature gas-cooled reactor (HTGR) source-term analyses. Each capsule in the AGR-3/4 experiment had four fuel compacts in the middle of two concentric rings of graphitic matrix material, PCEA graphite, or IG-110 graphite. In addition to the approximately 1898 tristructural isotropic (TRISO) coated particles in each compact, there were 20 designed to fail (DTF) particles coated only in pyrocarbon so that they released fission products into the surrounding cylindrical rings of carbonaceous materials. Destructive sampling of the rings involved machining/milling material from around the circumference of the rings, collecting that material, and performing radiochemical analyses on it. Milling operations were performed in multiple steps or segments, and each segment was generally 0.508 mm (0.020 in) thick. Knowing the radial position at which each segment was milled, the volume of the milled material at each segment, and the fission product content in each segment, the radial fission product concentrations were constructed for select isotopes in each ring. Ag-110m profiles had the most variation. Some profiles were peaked at an inner or outer surface. Some were peaked at the middle of the ring wall thickness. Some increased radially outward, and some decreased radially outward. These types of variations and the fact that the measured profiles do not generally compare favorably with the transport model employed for AGR-3/4 may adversely impact the ability to extract reasonable transport parameters for this isotope. In many cases, the Cs-134 profiles decreased somewhat linearly in the outward radial direction, and in cursory comparisons, the shapes of these profiles appeared similar to those from model predictions. The step changes in concentration across the inner-outer ring gap were generally consistent with the model predictions as well. In some cases, there were local maxima in concentration at the outer surface of the rings. This suggests that fission products could have transported in the small gaps between the inner ring and the outer ring and between the outer ring and the sink ring such that some portion of a given fission product can bypass diffusion through the ring itself. The analysis of the small nubs on the outer surfaces of some of the outer rings revealed fission product concentrations in the nubs that were often higher than in the outermost segments of the rings. This further supports the hypothesis that short-circuit, gap transport occurred, causing relatively high surface concentrations on the outer surfaces of the rings. Eu-154 and Sr-90 profiles tend to have very similar shapes, suggesting that they transport via the same mechanisms. The observed profiles were indicative of a transport process where the isotopes are sorbed on the inner surface of the ring, but diffusion into the ring from that surface is quite slow. Some elevated concentrations of Sr-90 (relative to Eu-154) on the outer surface of a ring suggested that rapid, gas-gap transport of gaseous precursor Kr-90 and volatile Rb-90 could have occurred prior to their decaying to Sr-90. Overall, the Eu 154 and Sr-90 profiles were still very similar, which indicates that the transport of short-lived Sr-90 precursors is not a major effect. In some capsules, the qualitative Sr-90 behavior across the ring gaps was consistent with the model (using the available legacy Sr-90 transport parameters), but in other capsules the model was inconsistent with the measurements and seems to underestimate the amount of Sr-90 in the outer rings. The total ring Sr-90 inventories were estimated for all the rings that were subject to physical sampling. These results will be used to adjust the predicted particle and/or compact releases used in the AGR-3/4 fission product transport model. Given the different irradiation temperatures among the capsules and the rings, it was not possible to discern fundamental differences in the transport of isotopes within the different carbon materials, i.e., graphitic matrix, IG-110, or PCEA. It may be possible to do this in the course of determining transport from the concentration profiles in future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Alpha/beta-gated gamma–gamma spectroscopy of mixed fission products for trace analysis

This experiment demonstrated that some hard-to-detect fission products can be quantified directly from mixed fission product samples using advanced radio-emission spectroscopy techniques. A gamma/gamma coincidence detector was combined with a custom liquid scintillation counter to evaluate the benefits of alpha and beta coincident and anti-coincident analysis of mixed fission product samples. Reduction factors in the minimum detectable activity (MDA) relative to traditional gamma counting were quantified for a set of radionuclides. For isotopes like 91Y, 144Ce, 147Nd, and 153Sm MDA reductions of factors of 1-2 were observed using optimized beta thresholds. Using beta anti-coincidence counting reductions of factors 3-5 were observed for isotopes like 89Sr, 95mNb, 115Cd, and 137Cs and factors of 5-7 for isotopes like 235U using alpha coincidence counting. The selectivity of gamma/gamma coincidence and reduction in backgrounds using beta gating were also demonstrated with 136Cs and 153Sm. The improvements observed in the detection sensitivity of fission products and actinides achieved by combining time-correlated radiation signatures is significant and of value to applications such as nuclear fission physics, environmental monitoring for nuclear waste, reactor accidental releases, and nuclear forensics.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗