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At least 19 records

Actinide, Activation Product and Fission Product Decay Data for Reactor-based Applications

The UK Activation Product Decay Data Library was first released in September 1977 as UK-PADD1, to be followed by regular improvements on an almost yearly basis up to the assembly of UKPADD6.12 in March 2013. Similarly, the UK Heavy Element and Actinide Decay Data Library followed in December 1981 as UKHEDD1, with the implementation of various modifications leading to UKHEDD2.6, February 2008. Both the data content and evaluation procedures are defined, and the most recent evaluations are described in terms of specific radionuclides and the resulting consistency of their recommended decay-data files. New versions of the UKPADD and UKHEDD libraries are regularly submitted to the NEA Data Bank for possible inclusion in the JEFF library.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Analysis of fission-product energy release after thermal-neutron fission of {sup 235}U and {sup 239}Pu by scale-6.1.3 with recently available fission product yield data - 14419

The fission-product energy release after the reactor shutdown has been an important issue from the perspective of the nuclear reactor safety in a very short time. The energy release from the decay of fission products can be estimated using a summation calculation method implemented in various burnup and depletion programs. The method generally depends on the nuclear structure data such as the decay and fission product yield data. In this study, the decay heat measurements after thermal-neutron fission of {sup 235}U and {sup 239}Pu have been evaluated by the ORIGEN-S with the decay and fission product yield libraries included in the SCALE-6.1.3 code package. In order to verify the importance of the fission product yield data when estimating the decay heats, new fission product yields libraries for ORIGEN-S have been generated based on the recently available fission product yield data such as ENDF/B-VII.1, JEFF-3.1.1, JENDL/FPY-2011, and JENDL-4.0. The new libraries were applied to the decay heat calculations, and the results were compared with those of the reference calculation based on the yield library that belongs to the SCALE-6.1.3. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Rapid activation product separations from fission products and soil matrixes

We report that a rapid method for the separation and qualitative analysis of several neutron activation products ( 198 Au, 192 Ir, 72 Ga, 51 Cr, 191/195m/197 Pt, 54 Mn, 57 Co, and 59 Fe) from fission products and soil matrixes has been developed. Analytes were isolated within 20 h using ion exchange chromatography. After separation, the activation products were characterized by γ-spectroscopy and inductively coupled plasma-optical emission spectroscopy. Validation experiments demonstrated versatility of the method, showing that the activation products could be isolated from fresh fission products and other contaminants associated with complex soil matrixes.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

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↗

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.↗

New Routes for the Production of Fission Products and Their Separation from the Source Material for the Development of Realistic Nuclear Forensic Debris

In order to fully characterize the NIF neutron spectrum, SAND-II-SNL software was requested/received from the Radiation Safety Information Computational Center. The software is designed to determine the neutron energy spectrum through analysis of experimental activation data. However, given that the source code was developed in Sparcstation 10, it is not compatible with current version of FORTRAN. Accounts have been established through the Lawrence Livermore National Laboratory’s High Performance Computing in order to access different compiles for FORTRAN (e.g. pgf77, pgf90). Additionally, several of the subroutines included in the SAND-II-SNL package have required debugging efforts to allow for proper compiling of the code.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Investigation of Inconsistent ENDF/B-VII.1 Independent and Cumulative Fission Product Yields with Proposed Revisions

A recent implementation of ENDF/B-VII.1 independent fission product yields and nuclear decay data identified inconsistencies in the data caused by the use of updated nuclear schemes in the decay sub-library that are not reflected in legacy fission product yield data. Recent changes in the decay data sub-library, particularly the delayed neutron branching fractions, result in calculated fission product concentrations that do not agree with the cumulative fission yields in the library as well as with experimental measurements. To address these issues, a comprehensive set of independent fission product yields was generated for thermal and fission spectrum neutron-induced fission for {sup 235,238}U and {sup 239,241}Pu in order to provide a preliminary assessment of the updated fission product yield data consistency. These updated independent fission product yields were utilized in the ORIGEN code to compare the calculated fission product inventories with experimentally measured inventories, with particular attention given to the noble gases. Another important outcome of this work is the development of fission product yield covariance data necessary for fission product uncertainty quantification. The evaluation methodology combines a sequential Bayesian method to guarantee consistency between independent and cumulative yields along with the physical constraints on the independent yields. This work was motivated to improve the performance of the ENDF/B-VII.1 library for stable and long-lived fission products. The revised fission product yields and the new covariance data are proposed as a revision to the fission yield data currently in ENDF/B-VII.1.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Identifying and quantifying short-lived fission products from thermal fission of HEU using portable HPGe detectors

Due to the emerging potential for trafficking of special nuclear material, research programs are investigating current capabilities of commercially available portable gamma ray detection systems. Presented in this paper are the results of three different portable high-purity germanium (HPGe) detectors used to identify short-lived fission products generated from thermal neutron interrogation of small samples of highly enriched uranium. Samples were irradiated at the Washington State University (WSU) Nuclear Radiation Center’s 1MW TRIGA reactor. The three portable, HPGe detectors used were the ORTEC MicroDetective, the ORTEC Detective, and the Canberra Falcon. Canberra’s GENIE-2000 software was used to analyze the spectral data collected from each detector. Ultimately, these three portable detectors were able to identify a large range of fission products showing potential for material discrimination.

detectors↗

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.

Pierson, Bruce D.↗

Analysis of fission product distribution and composition in the TRISO layers of AGR-2 fuel

The Advanced Gas Reactor Fuel Qualification and Development Program's second irradiation experiment focuses on the performance of fuel particles produced in an industrial-scale coating system. Ongoing post-irradiation examination efforts include exploration of tristructural-isotropic (TRISO)-coated particle fuel in the as-irradiated state and subjected to elevated temperatures to simulate accident conditions (safety testing). Scanning electron microscopy (SEM) analysis with energy dispersive x-ray spectroscopy (EDS) was used to show the distribution of fission products within the TRISO layers. The SEM analysis showed a variation of fission product distribution within the TRISO layers based on irradiation temperature, burnup, individual particle fission product inventory retention, safety testing history and kernel composition. Finally, a key observation is that particles which likely experienced higher temperature showed a greater diversity and distribution of fission products and actinides in the IPyC and SiC layers, implying that temperature is a dominant variable influencing the transport of radionuclides in TRISO particles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Electron Microscopic Evaluation and Fission Product Identification of Irradiated TRISO Coated Particles from the AGR-1 Experiment: A Preliminary Review

Post-irradiation examination of coated particle fuel from the AGR-1 experiment is in progress at Idaho National Laboratory and Oak Ridge National Laboratory. In this paper a brief summary of results from characterization of microstructures in the coating layers of selected irradiated fuel particles with burnup of 11.3% and 19.3% FIMA will be given. The main objectives of the characterization were to study irradiation effects, fuel kernel porosity, layer debonding, layer degradation or corrosion, fission-product precipitation, grain sizes, and transport of fission products from the kernels across the TRISO layers. Characterization techniques such as scanning electron microscopy, transmission electron microscopy, energy dispersive spectroscopy, and wavelength dispersive spectroscopy were used. A new approach to microscopic quantification of fission-product precipitates is also briefly demonstrated. Microstructural characterization focused on fission-product precipitates in the SiC-IPyC interface, the SiC layer and the fuel-buffer interlayer. The results provide significant new insights into mechanisms of fission-product transport. Although Pd-rich precipitates were identified at the SiC-IPyC interlayer, no significant SiC-layer thinning was observed for the particles investigated. Characterization of these precipitates highlighted the difficulty of measuring low concentrations of Ag in precipitates with significantly higher concentrations of Pd and U. Different approaches to resolving this problem are discussed. An initial hypothesis is provided to explain fission-product precipitate compositions and locations. No SiC phase transformations were observed and no debonding of the SiC-IPyC interlayer as a result of irradiation was observed for the samples investigated. Lessons learned from the post-irradiation examination are described and future actions are recommended.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Fission Product Measurement and Off-Gas Scrubbing for Molten Salt Reactors

The Advanced Reactor Technology (ART) program within the US Department of Energy is developing understanding and technology to help facilitate molten salt reactor commercialization. The off-gas system for MSRs plays an important role in fission product confinement. The strategy used in MSRs is different from that used in pressurized water reactors, where the off-gas mainly mitigates severe accident conditions. Although the details vary depending on the reactor type, the off-gas system of MSRs must reduce radiological emissions to the environment to levels comparable with those mandated for non-MSRs. Systems with molten fuel salt will accumulate material in the headspace: volatile fission products, entrained salt particles, graphite fines, and material that has low solubility in the salt. Operators of thermal fluoride salt reactors with encapsulated fuel will also be concerned about tritium generation and migration. The fission and activation products include noble gases, their daughters, tritium, and aerosolized salt constituents. Their concentrations will generally increase with reactor burnup, depending on the various strategies used for mitigation. Additionally, tritium capture and removal has been reviewed by several groups.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Fission product release and survivability of UN-kernel LWR TRISO fuel

A thermomechanical assessment of the LWR application of TRISO fuel with UN kernels was performed. Fission product release under operational and transient temperature conditions was determined by extrapolation from fission product recoil calculations and limited data from irradiated UN pellets. Both fission recoil and diffusive release were considered and internal particle pressures computed for both 650 and 800 um diameter kernels as a function of buffer layer thickness. These pressures were used in conjunction with a finite element program to compute the radial and tangential stresses generated within a TRISO particle undergoing burnup. Creep and swelling of the inner and outer pyrolytic carbon layers were included in the analyses. A measure of reliability of the TRISO particle was obtained by computing the probability of survival of the SiC barrier layer and the maximum tensile stress generated in the pyrolytic carbon layers from internal pressure and thermomechanics of the layers. These reliability estimates were obtained as functions of the kernel diameter, buffer layer thickness, and pyrolytic carbon layer thickness. The value of the probability of survival at the end of irradiation was inversely proportional to the maximum pressure.

99 GENERAL AND MISCELLANEOUS↗

Energy Dependence of Fission Product Yields from {sup 235}U, {sup 238}U and {sup 239}Pu for Incident Neutron Energies Between 0.5 and 14.8 MeV

Fission Product Yields (FPY) have historically been one of the most observable features of the fission process. They are known to have strong variations that are dependent on the fissioning species, the excitation energy, and the angular momentum of the compound system. However, consistent and systematic studies of the variation of these FPY with energy have proved challenging. This is caused primarily by the nature of the experiments that have traditionally relied on radiochemical procedures to isolate specific fission products. Although radiochemical procedures exist that can isolate all products, each element presents specific challenges and introduces varying degrees of systematic errors that can make inter-comparison of FPY uncertain. Although of high importance in fields such as nuclear forensics and Stockpile Stewardship, accurate information about the energy dependence of neutron induced FPY are sparse, due primarily to the lack of suitable monoenergetic neutron sources. There is a clear need for improved data, and to address this issue, a collaboration was formed between Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL) and the Triangle Universities Nuclear Laboratory (TUNL) to measure the energy dependence of FPY for {sup 235}U, {sup 238}U and {sup 239}Pu. The measurements have been performed at TUNL, using a 10 MV Tandem Van de Graaff accelerator to produce monoenergetic neutrons at energies between 0.6 MeV to 14.8 MeV through a variety of reactions. The measurements have utilized a dual-fission chamber, with thin (10-100 μg/cm2) reference foils of similar material to a thick (100-400 mg) activation target held in the center between the chambers. This method allows for the accurate determination of the number of fissions that occurred in the thick target without requiring knowledge of the fission cross section or neutron fluence on target. Following activation, the thick target was removed from the dual-fission chamber and gamma-ray counted using shielded HPGe detectors for a period of 1-2 months to determine the yield of various fission products. To the extent possible all irradiation and counting procedures were kept the same to minimize sources of systematic errors. FPY have been determined at incident neutron energies of 0.6, 1.4, 2.4, 3.5, 4.6, 5.5, 8.9 and 14.8 MeV.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Waste form evaluation for RECl 3 and REO x fission products separated from used electrochemical salt

The work presented here is based off the concept that the rare earth chloride (RECl3) fission products mixture within the used electrorefiner (ER) salt can be selectively removed as RECl3 (not yet demonstrated) or precipitated out as REOCl through oxygen sparging (has been demonstrated). This paper presents data showing the feasibility of immobilizing a mixture of RECl3’s at 10 mass% into a TeO2-PbO glass and it shows that this same mixture of RECl3’s can be oxidized to REOCl at 300°C and then to REOx by 1200°C. When the REOx mixture is heated at temperatures >1200°C, the ratios of REOx’s change. The mixture of REOx was then immobilized in a LABS glass at a high loading of 60 mass%. Both the TeO2-PbO glass and LABS glass systems show good chemical durability. The advantages and disadvantages of tellurite and LABS glasses are compared.

Riley, Brian J. (ORCID:0000000277456730)↗

Cement As a Waste Form for Nuclear Fission Products: The Case of 90 Sr and Its Daughters [Cement As a Container for Nuclear Fission Products: The Case of 90 Sr and Its Daughters]

One of the main challenges faced by the nuclear industry is the long-term confinement of nuclear waste. Because it is inexpensive and easy to manufacture, cement is the material of choice to store large volumes of radioactive materials, in particular the low-level medium-lived fission products. It is therefore of utmost importance to assess the chemical and structural stability of cement containing radioactive species. Here, we use ab initio calculations based on density functional theory (DFT) to study the effects of 90 Sr insertion and decay in C–S–H (calcium-silicate-hydrate) in order to test the ability of cement to trap and hold this radioactive fission product and to investigate the consequences of its β-decay on the cement paste structure. We show that 90 Sr is stable when it substitutes the Ca 2+ ions in C–S–H, and so is its daughter nucleus 90 Y after β-decay. Interestingly, 90 Zr, daughter of 90 Y and final product in the decay sequence, is found to be unstable compared to the bulk phase of the element at zero K but stable when compared to the solvated ion in water. Furthermore, cement appears as a suitable waste form for 90 Sr storage.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗