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At least 217 records · Page 12

Materials Data on Pu(FeSi)2 by Materials Project

PuFe2Si2 crystallizes in the tetragonal I4/mmm space group. The structure is three-dimensional. Pu3+ is bonded in a distorted body-centered cubic geometry to eight equivalent Si4- atoms. All Pu–Si bond lengths are 3.02 Å. Fe+2.50+ is bonded to four equivalent Si4- atoms to form a mixture of corner and edge-sharing FeSi4 tetrahedra. All Fe–Si bond lengths are 2.27 Å. Si4- is bonded in a 9-coordinate geometry to four equivalent Pu3+, four equivalent Fe+2.50+, and one Si4- atom. The Si–Si bond length is 2.42 Å.

36 MATERIALS SCIENCE↗

Materials Data on Pu(SiOs)2 by Materials Project

PuOs2Si2 crystallizes in the tetragonal I4/mmm space group. The structure is three-dimensional. Pu4+ is bonded in a 8-coordinate geometry to eight equivalent Os2- atoms. All Pu–Os bond lengths are 3.18 Å. Os2- is bonded in a 4-coordinate geometry to four equivalent Pu4+ and four equivalent Si atoms. All Os–Si bond lengths are 2.40 Å. Si is bonded in a 5-coordinate geometry to four equivalent Os2- and one Si atom. The Si–Si bond length is 2.38 Å.

36 MATERIALS SCIENCE↗

Materials Data on Pu(CoSi)2 by Materials Project

PuCo2Si2 crystallizes in the tetragonal I4/mmm space group. The structure is three-dimensional. Pu4+ is bonded in a body-centered cubic geometry to eight equivalent Si4- atoms. All Pu–Si bond lengths are 3.01 Å. Co2+ is bonded to four equivalent Si4- atoms to form a mixture of corner and edge-sharing CoSi4 tetrahedra. All Co–Si bond lengths are 2.28 Å. Si4- is bonded in a 9-coordinate geometry to four equivalent Pu4+, four equivalent Co2+, and one Si4- atom. The Si–Si bond length is 2.52 Å.

36 MATERIALS SCIENCE↗

Materials Data on Pu(SiRu)2 by Materials Project

PuRu2Si2 crystallizes in the tetragonal I4/mmm space group. The structure is three-dimensional. Pu3+ is bonded in a distorted body-centered cubic geometry to eight equivalent Si4- atoms. All Pu–Si bond lengths are 3.20 Å. Ru+2.50+ is bonded to four equivalent Si4- atoms to form a mixture of corner and edge-sharing RuSi4 tetrahedra. All Ru–Si bond lengths are 2.39 Å. Si4- is bonded in a 9-coordinate geometry to four equivalent Pu3+ and four equivalent Ru+2.50+ atoms.

36 MATERIALS SCIENCE↗

Materials Data on Pu(MnSi)2 by Materials Project

PuMn2Si2 crystallizes in the tetragonal I4/mmm space group. The structure is three-dimensional. Pu4+ is bonded in a distorted body-centered cubic geometry to eight equivalent Si4- atoms. All Pu–Si bond lengths are 2.96 Å. Mn2+ is bonded to four equivalent Si4- atoms to form a mixture of edge and corner-sharing MnSi4 tetrahedra. All Mn–Si bond lengths are 2.36 Å. Si4- is bonded in a 9-coordinate geometry to four equivalent Pu4+, four equivalent Mn2+, and one Si4- atom. The Si–Si bond length is 2.37 Å.

36 MATERIALS SCIENCE↗

Materials Data on Pu(SiPt)2 by Materials Project

PuPt2Si2 crystallizes in the tetragonal P4/nmm space group. The structure is three-dimensional. Pu4+ is bonded in a 8-coordinate geometry to eight Pt2- atoms. There are four shorter (3.20 Å) and four longer (3.28 Å) Pu–Pt bond lengths. There are two inequivalent Pt2- sites. In the first Pt2- site, Pt2- is bonded to four equivalent Pu4+ and four equivalent Si atoms to form a mixture of distorted edge and face-sharing PtPu4Si4 tetrahedra. All Pt–Si bond lengths are 2.49 Å. In the second Pt2- site, Pt2- is bonded in a 5-coordinate geometry to four equivalent Pu4+ and five Si atoms. There are one shorter (2.38 Å) and four longer (2.43 Å) Pt–Si bond lengths. There are two inequivalent Si sites. In the first Si site, Si is bonded in a 5-coordinate geometry to five Pt2- atoms. In the second Si site, Si is bonded to four equivalent Pt2- atoms to form a mixture of edge and corner-sharing SiPt4 tetrahedra.

36 MATERIALS SCIENCE↗

High-precision measurement of U-Pu-Np-Am concentrations and isotope ratios in environmental reference materials by mass spectrometry

Herein, we report results of precise and sensitive mass spectrometric measurements of uranium, plutonium, neptunium, and americium concentrations and isotope ratios in a variety of environmental reference materials. Most of our work has been done on NIST SRM 4350b, River Sediment, but we also present results for NIST SRM 4354, Lake Sediment; NIST SRMs 4355 and 4355a, Peruvian Soil; NIST SRM 4357, Ocean Sediment; NIST SRM 1648a, Urban Particulate Material; NIST SRM 1649b, Urban Dust; IAEA CRM 385, Ocean Sediment; USGS BCR-2, Columbia River Basalt; and USGS BHVO-2, Hawaiian Volcanic Observatory Basalt. These materials reflect a wide range in long-lived actinide concentrations (e.g. 1E4 to 1E10 atoms 239 Pu/g) and isotope ratios. Measurements were performed in a clean laboratory by isotope dilution, multi-collector thermal ionization and multi-collector inductively coupled plasma mass spectrometry. In general, our results are in agreement with, but lower the uncertainty of, literature or certificate values for these reference materials. Our uranium results for the basalts also confirm previously reported high-precision mass spectrometric results from our laboratory. In many cases our measurements of U-Pu-Np-Am nuclides appear to be novel.

58 GEOSCIENCES↗

Cross-Code Verification of Neutronics Analysis Tools at INL Applied for 238 Pu Production in the Advanced Test Reactor

Here, analyses are completed for experiments prior to experiment irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Various codes are used to qualify all experiments planned for insertion in the reactor, thereby ensuring that all safety and programmatic requirements are satisfied preirradiation. Among the common experiment analysis tools at INL are MCNP5 coupled to ORIGEN2 (MOPY) and MC21. MOPY uses MCNP5 for transport calculations along with calculations for fluxes and select reaction rates, and then ORIGEN2 handles the step-by-step and postirradiation depletion. MC21 handles all in-reactor transport and step-by-step, during-irradiation, depletion calculations, and then ORIGEN (SCALE 6.2.3) is used for decay and dose calculations postirradiation. The MOPY results, along with those obtained via two variations of the MC21 model, were compared in terms of 238 Pu production in the ATR’s H10 position. For the MOPY model, the MC21 model utilizing the HELIOS-based fission product (FP) library, and the MC21 model utilizing the expanded 1300 FP library, the during-cycle irradiation in-core heating results were sufficiently equivalent; however, the MOPY model and the MC21 model with the HELIOS library showed some differences relating to the respective FP libraries. Ultimately, the MC21 model with a 1300 FP library produced the most consistent results throughout the cycle, whereas the MC21 model that utilized the (smaller) HELIOS library was able to handle during-irradiation analysis but lacked certain short-lived FPs that significantly contributed to the total decay heat at shutdown. MOPY, on the other hand, was found to overpredict fission gas production, as a result of limitations in the ORIGEN2 code.

ATR↗

Neutronic Safety Analysis of Pu-238 Production at Idaho National Laboratory

This analysis was completed to support the irradiation of plutonium fuel services (PFS) targets in the NEFT in the Advanced Test Reactor (ATR) as a part of the campaign to restart domestic production of plutonium-238 used in radioisotope power systems (RPS) by the National Aeronautical and Space Administration (NASA) and Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Nuclear Infrastructure Program (NE-3). Referred to as the PFS-ATR-GEN1-NEFT experiment, the assembly was designed to hold 46 PFS targets in the NEFT. The scope of this paper is to outline the MOPY method used to calculate the heat generation rates (HGRs), flux, fission density, and quantify the viability of the target design for Pu-238 production in ATR.

07 ISOTOPE AND RADIATION SOURCES↗

Evaluation of Pu Solubility in Glass for Sludge Batch 11

Stainless-steel clad Pu from Japan’s Fast Critical Assembly (FCA) reactor is currently being dispositioned at the Savannah River Site. The electrolytic dissolver, operated by Savannah River Nuclear Solutions in H-Canyon, is being utilized to dissolve the material. The resulting solutions are transferred to the Concentration, Storage, and Transfer Facilities, operated by Savannah River Mission Completion (SRMC), for subsequent vitrification at the Defense Waste Processing Facility (DWPF). In support of the FCA mission startup, a preliminary evaluation was conducted by the Savannah River National Laboratory to assess the impact of the FCA discards on the liquid waste system.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Pu-238 Production Calculator and Estimation Tool in the Advanced Test Reactor

A MS Excel tool was created to help track Pu-238 production, Np-237 feedstock use, estimate production scenarios, and coordinate planning for shipments at the Idaho National Laboratory (INL) and Advanced Test Reactor (ATR). As-run estimates were used as a basis for beginning and end of life isotopic concentrations. Multiple target types were created to demonstrate a proof of concept while actual results were not yet available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Pu-238 Production At Idaho National Laboratory

Overview of INL's work on Pu-238 production presented to MFC Strategic Advisory Council (SAC). Discusses program status, upcoming work, and future planned activities.

07 ISOTOPE AND RADIATION SOURCES↗

Thermochemically-informed mass transport model for interdiffusion of U and Zr in irradiated U-Pu-Zr fuel with fission products

A new formulation for redistribution of constituents in metallic U-Pu-Zr nuclear fuel is presented that can incorporate the contributions from fission products. The formulation is based on the thermodynamic driving forces derived from the generalized chemical potential that includes effects of composition and temperature. As a result, the redistribution model can readily account for the composition changes due to the generation of fission products while using only a limited set of transport coefficients. The thermodynamic model for the metallic fuel and thermochemistry solver Thermochimica were coupled with the nuclear fuel performance code BISON to implement the redistribution model. The simulations reproduce an experimentally observed Zr-depletion zone in the mid-radius region of an irradiated fuel slug. The generation of fission products during burnup is shown to have a stabilizing effect on the fuel chemistry, slowing the rate of U-Zr interdiffusion and reducing the size of the Zr-depleted zone. Here, the overall objective of the work is to advance predictive capabilities of fission product behaviour in U-Zr metallic fuel in the context of fuel performance and safety.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Evaluation of the Prompt Fission Gamma Properties for Neutron Induced Fission of 235,238 U and 239 Pu

Herein we present details of the prompt fission gamma property evaluations included in the released ENDF/B-VIII.0 files. The average prompt gamma multiplicity, the average prompt gamma spectrum, the average total prompt gamma energy released and the prompt gamma multiplicity distributions, included in the ENDF/B-VIII.0 files, are presented for 235 U(n,f), 238 U(n,f), and 239 Pu(n,f) reactions, as a function of incident neutron energy up to 30 MeV. The evaluation is based on available experimental data and model calculations. Notable data include recent measurements by experimental groups at Los Alamos and Lawrence Livermore National Laboratories, at Geel and Budapest, at CEA labs, as well as seminal historical gamma production measurements by Drake at Los Alamos.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Comparison of Results from Recent NNSA and CEA Measurements of the 239 Pu(n, f) Prompt Fission Neutron Spectrum

The National Nuclear Security Administration (NNSA)/DP French Alternative Energies and Atomic Energy Commission (CEA)/DAM agreement on cooperation on fundamental science is a U.S.-French collaborative effort to combine intellectual and experimental resources and further the relevant nuclear science. Recently, both the NNSA and CEA experimental teams performed high-statistics measurements of the 239 Pu(n, f) prompt fission neutron spectrum (PFNS) at the Los Alamos Neutron Science Center, both of which were recently published in the journal Physical Review C. These separate measurements used the same experimental area and a common neutron detector array, but differ in many aspects, including background assessments, data acquisition systems and philosophies, fission detectors, and PFNS extraction techniques. Hence, some aspects of the experimental methods and associated uncertainties are highly correlated while others are independent. The results from both measurements broke new ground for PFNS measurements given their higher accuracy and more detailed study of corrections necessary for the measured quantity compared to existing literature measurements, and both will significantly impact PFNS nuclear data evaluations for the foreseeable future. Here, the focus of this work is to document a comparison of the results from these distinct measurements in terms of the acquired data, the PFNS results, and the measured average PFNS energies. While systematic differences between the PFNS results are present on the 1–3% level, the acquired data relative to each respective measurement at low incident neutron energies are in remarkable agreement, as are the conclusions regarding the magnitude and position of features in the PFNS relating to second-chance fission, third-chance fission, and pre-equilibrium neutron emission.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Measurement of the 239 Pu(n,f)/ 235 U(n,f) Cross-Section Ratio with the NIFFTE fission Time Projection Chamber

In this work, the 239 Pu(n,f)/ 235 U(n,f) cross-section ratio has been measured with the fission Time Projection Chamber (fissionTPC) from 100 keV to 100 MeV. The fissionTPC provides three-dimensional reconstruction of fission-fragment ionization profiles, allowing for a precise quantification of measurement uncertainties. The measurement was performed at the Los Alamos Neutron Science Center which provides a pulsed white source of neutrons. The data are recommended to be used as a cross-section ratio shape. A discussion of the status of the absolute normalization and comparisons to ENDF evaluations and previous measurements is included.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Beta-delayed gamma spectra compilation and analysis following the thermal-neutron induced fission of 235 U, 239,241 Pu

The integral gamma and electron spectra emitted by fission products, also known as delayed gamma and electron spectra, were measured at Oak Ridge National Laboratory in the 1970s for the thermal-neutron induced fission of 235 U and 239,241 Pu. Scintillator detectors were used to measure these spectra, data used later on to obtain decay heat values - that is, the spectra mean values per unit time as function of time - work that was published in the Nuclear Science and Technology journal; the spectral data, however, was only published in laboratory reports. Here, in this work, we analyze the gamma spectra data using modern methods and nuclear databases to reveal the signature of individual fission products as well as to gauge the performance of the ENDF/B-VIII.0 decay data sub-library, concluding about possible future enhancements in predictive capabilities.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗