Results for “Pu” — Page 44

DOE OSTI2022

Testing nuclear data libraries with burnup for reactor applications

Recently US and European nuclear data libraries have been released, namely ENDF/B-VIII.0 and JEFF-3.3 libraries, which are the result of years of evaluation and validation work in both communities. Consequently, efforts have been made to validate the evaluations in calculations of integral experiments (critical benchmarks, ..etc). Nevertheless, less stringent testing and validation efforts were performed on burnup applications before releasing the libraries. The presented work focusses on the testing of these recent nuclear data libraries for burnup calculations on two benchmarks at the pin and at the assembly level. Monte-Carlo depletion calculations were performed using the VESTA 2.2 code. The K{sub ∞} results between nuclear data libraries are compared. A strong k{sub ∞} bias is observed with burnup using both JEFF-3.3 and JEFF-4T0 compared to all other libraries, and especially ENDF/B-VIII.0, consisting in a strong k{sub ∞} over-estimation at low burnup and a high under-estimation at high burnup. JEFF-3.3 {sup 235}U and {sup 239}Pu evaluations mainly explain this result, as well as fission yields. New {sup 235}U and {sup 239}Pu evaluations were proposed for JEFF-4T0, but they do not address totally the bias issue, even if JEFF-4T0 {sup 239}Pu allows slightly reducing the k{sub ∞} underestimation at high burnup.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Pre-Transient Characterization of Historic EBR-II Pins for Transient Testing

Current interest in sodium-cooled fast reactor (SFR) designs, such as TerraPower’s Natrium Reactor, has highlighted the need for advanced reactor fuel technology development. Modern U-Zr and U- Pu-Zr pin designs are primary candidates to fuel SFRs and boast high fuel utilization capacity, increased fuel-cladding compatibility, and improved safety through inherent feedback mechanisms. Despite over 60 years of metallic fuel irradiation, uncertainties exist in the performance of the fuel system, particularly under transient overpower (TOP) and loss of flow (LOF) scenarios. Throughout historical testing within the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF), fuel behavior has demonstrated benign response to transient reactor conditions; however, accurate predictions of failure thresholds to inform operational limitations rely heavily on fuel composition, burnup, and irradiation history. In expanding TOP and LOF testing, the Transient Heat sink Overpower Response (THOR) Capsule will be used to test modern fuel technologies in a static sodium environment in the Transient Reactor Test (TREAT) Facility. The THOR capsule is highly instrumented and will provide time-dependent thermal behavior of SFR fuel pins subjected to accident conditions within TREAT. The THOR-Metallic (THOR- M) campaign aims to validate and expand historical TOP and LOF testing on high burnup U-Zr and U-Pu- Zr fuel alloys previously irradiated in EBR-II by running the rods to failure. This contribution focuses primarily on the pre-transient engineering-scale destructive and non- destructive characterization that has been conducted on both the test and sibling pins used for the TOP and LOF tests. All pins underwent visual examination, neutron radiography, element contact profilometry, and precise gamma scan. The sibling pins used for each test were further analyzed using gas assay, sampling, and recharge analysis (GASR), and optical microscopy. The results from each technique confirmed that the fuel pins were intact and devoid of any atypical developments when compared to historical data. Additionally, the analyzed measurements establish a baseline for comparison to post-transient analysis. Key fuel behaviors quanitifed include axial elongation of the fuel column, diametral strain of the pin, patterns in fluff structure geometry, changes in axial isotope distribution, evolution of constituent redistribution, porosity, and fission gas release. The pre-transient measurements and changes attributed to transient behavior from post-transient measurement will be compared to historical data to capture the behavioral dependence on composition, burnup, and irradiation history. Results from this work advance the initiatives of the THOR-M campaign, which aid in informing fuel performance models and establishing safety criteria for SFR operational limits. The novel combination of test environment, in-situ instrumentation, and comprehensive suite of characterization methods provides greater understanding of transient fuel behavior. Overall, information on the time and condition of pin failure for high burnup U-Pu-Zr will greatly expand the limited existing TOP and LOF test data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Pre-Transient Characterization of Historic EBR-II Pins for Transient Testing

Current interest in sodium-cooled fast reactor (SFR) designs, such as TerraPower’s Natrium Reactor, has highlighted the need for advanced reactor fuel technology development. Modern U-Zr and U- Pu-Zr pin designs are primary candidates to fuel SFRs and boast high fuel utilization capacity, increased fuel-cladding compatibility, and improved safety through inherent feedback mechanisms. Despite over 60 years of metallic fuel irradiation, uncertainties exist in the performance of the fuel system, particularly under transient overpower (TOP) and loss of flow (LOF) scenarios. Throughout historical testing within the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF), fuel behavior has demonstrated benign response to transient reactor conditions; however, accurate predictions of failure thresholds to inform operational limitations rely heavily on fuel composition, burnup, and irradiation history. In expanding TOP and LOF testing, the Transient Heat sink Overpower Response (THOR) Capsule will be used to test modern fuel technologies in a static sodium environment in the Transient Reactor Test (TREAT) Facility. The THOR capsule is highly instrumented and will provide time-dependent thermal behavior of SFR fuel pins subjected to accident conditions within TREAT. The THOR-Metallic (THOR- M) campaign aims to validate and expand historical TOP and LOF testing on high burnup U-Zr and U-Pu- Zr fuel alloys previously irradiated in EBR-II by running the rods to failure. This contribution focuses primarily on the pre-transient engineering-scale destructive and non- destructive characterization that has been conducted on both the test and sibling pins used for the TOP and LOF tests. All pins underwent visual examination, neutron radiography, element contact profilometry, and precise gamma scan. The sibling pins used for each test were further analyzed using gas assay, sampling, and recharge analysis (GASR), and optical microscopy. The results from each technique confirmed that the fuel pins were intact and devoid of any atypical developments when compared to historical data. Additionally, the analyzed measurements establish a baseline for comparison to post-transient analysis. Key fuel behaviors quanitifed include axial elongation of the fuel column, diametral strain of the pin, patterns in fluff structure geometry, changes in axial isotope distribution, evolution of constituent redistribution, porosity, and fission gas release. The pre-transient measurements and changes attributed to transient behavior from post-transient measurement will be compared to historical data to capture the behavioral dependence on composition, burnup, and irradiation history. Results from this work advance the initiatives of the THOR-M campaign, which aid in informing fuel performance models and establishing safety criteria for SFR operational limits. The novel combination of test environment, in-situ instrumentation, and comprehensive suite of characterization methods provides greater understanding of transient fuel behavior. Overall, information on the time and condition of pin failure for high burnup U-Pu-Zr will greatly expand the limited existing TOP and LOF test data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Purification Techniques for Actinide Radiolysis Studies

Our fundamental understanding of actinide radiation-induced redox chemistry is crucial due to their unavoidable exposure to ionizing radiation fields, both inherent and from in-process applications. Plutonium (Pu) and americium (Am) both possess multiple oxidation states, the careful manipulation of which are essential in the study and utilization of their rich chemistry, developing new technologies, and securing the long-term sustainability of nuclear power. However, knowledge in this area is far from complete. We have studied the radiation-induced chemistry of both Pu and Am through a variety of techniques, including gamma irradiation, in-situ alpha irradiation and pulse radiolysis experiments. However, for the collection of accurate data, thorough purification and quantification of actinide-containing solutions is required. This presentation will cover the purification and quantification techniques employed for the radiolysis experiments described in our recent publications: Kynman et al., Multiscale Modeling of Plutonium Radiation Chemistry in Nitric Acid Solutions. 1. Cobalt-60 Gamma Irradiation of Pu(IV) [https://pubs.acs.org/doi/10.1021/acs.inorgchem.4c00138] and Kynman et al., Generation and Study of Am(IV) by Temperature-Controlled Electron Pulse Radiolysis [https://doi.org/10.1039/D4DT00991F].

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2025

Investigating Radiation-Induced Actinide Species in Solution

Our fundamental understanding of actinide radiation-induced redox chemistry is crucial to nuclear fuel cycle development, due to the unavoidable exposure of these elements to ionizing radiation fields, both inherent and from in-process applications. Plutonium (Pu) and americium (Am) both possess multiple oxidation states, the careful manipulation of which are essential in the study and utilization of their rich chemistry, developing new nuclear technologies, and securing the long-term sustainability of nuclear power. However, knowledge in this area is far from complete. Consequently, we have studied the radiation-induced chemistry of both Pu and Am through a variety of techniques. Temperature-controlled electron pulse radiolysis has been used to study Am for the first time, determining the feasibility of Am redox reactions under used nuclear fuel reprocessing conditions, (e.g. nitric acid, non-ambient temperature). Additionally, we have developed an experimentally evaluated multi-scale computer model for the prediction of gamma radiation-induced Pu redox chemistry due to radiolysis and disproportionation reactions in concentrated nitric acid solutions.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2025

Investigating Radiation-Induced Actinide Species in Solution

Our fundamental understanding of actinide radiation-induced redox chemistry is crucial due to their unavoidable exposure to ionizing radiation fields, both inherent and from in-process applications. Plutonium (Pu) and americium (Am) both possess multiple oxidation states, the careful manipulation of which are essential in the study and utilization of their rich chemistry, developing new nuclear technologies, and securing the long-term sustainability of nuclear power. However, knowledge in this area is far from complete. Consequently, we have studied the radiation-induced chemistry of both Pu and Am through a variety of techniques. Temperature-controlled electron pulse radiolysis has been used to study Am for the first time, determining the feasibility of Am redox reactions under used nuclear fuel reprocessing conditions, (e.g. nitric acid, non-ambient temperature). Additionally, we developed an experimentally evaluated multi-scale computer model for the prediction of gamma radiation-induced Pu redox chemistry due to radiolysis and disproportionation reactions in concentrated nitric acid solutions.

actinide↗

DOE OSTI2023

Development of the generation III (ATR generation I) Plutonium-238 production target design

A domestic supply chain to produce 238 Pu fuel for radioisotope thermoelectric generators is critical for enabling future space exploration. A multi-laboratory effort has worked to establish a common target design to efficiently produce 238 Pu in two research reactors. This approach ensures that the annual production goals set forth by NASA are met, while also establishing redundant production capabilities. Here, this paper describes the effort to develop the common target design as well as considerations for future applications for the irradiation platform.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Preparation and characterization of aminated co-solvent enhanced lignocellulosic fractionation lignin as a renewable building block for the synthesis of non-isocyanate polyurethanes

Lignin is a major bioresource from 2nd generation biorefinery process streams and from pulp and paper production that is still severely underutilized nowadays. Its natural polymeric and aromatic structures make it suitable as a potential building block for biopolymer production. Due to the growing environmental and health concerns over conventional polyurethane (PU) manufacturing processes that utilize toxic isocyanates, there is a great interest to develop a non-isocyanate PU (NIPU) route. Herein, we demonstrate a novel strategy that uses highly pure lignin isolated from co-solvent enhanced lignocellulosic fractionation (CELF) pretreatment of poplar wood to produce biobased NIPUs. In this strategy, hardwood poplar is firstly fractionated via a CELF pretreatment to produce a clean lignin stream that is rich in phenolics. The CELF lignin was then aminated by a Mannich reaction, and the aminated CELF lignin was finally reacted with bicyclic carbonates to yield an advanced NIPU. In this work, the structure, mechanical, and thermal properties of these novel NIPUs were elucidated by FTIR, NMR, tensile test, dynamic mechanical analysis (DMA), and thermogravimetric analysis (TGA). The mechanical properties of these lignin based NIPUs could be tuned from highly rigid to elastic polymers by simply modifying the lignin constituents of the polymer. The thermal stabilities of NIPUs were improved because of the addition of aminated lignin, and NIPU containing 55 and 23 wt% lignin showed the highest tensile strength (~1.2 MPa) and elongation at break (~140%), respectively. The results obtained in this study demonstrate that the reaction of cyclic carbonate with aminated lignin represents an interesting pathway for the synthesizing of lignin based NIPU with a relative high lignin content.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2022

Transmission Electron Microscopy based Characterization of a U-20Pu-10Zr Fuel Irradiated in Experimental Breeder Reactor-II

U-Pu-Zr metallic fuels are important fuel candidate for future advanced and/or test reactors. To better understand fuel performance and guide future fuel development, a U-20Pu-10Zr (in weight) metallic fuel irradiated in the Experimental Breeder Reactor was revisited using advanced electron microscopies. This fuel was irradiated to a burnup of 7.6 % at a peak cladding temperature of 521 °C. In this research, several transmission electron microscopy samples were extracted from different but representative radial locations of the fuel cross section using focused ion beam technique. Phase identification and chemical analysis with sub-micron spatial resolution was carried out using scanning transmission electron microscopy. Multiple phenomena that are critical to fuel performance, such as Zr phase and its distribution in a different matrix, were revealed in an unpreceded and highly detailed manner. An improved understanding of fuel reconstruction in U-Pu-Zr metallic fuel under reactor irradiation is provided.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Life-cycle analysis of microalgae-based polyurethane foams

Polyurethane plastics are essential in many consumer and commercial products such as insulation, furniture, automotive interiors, and clothing. Pathways for producing polyurethane from microalgae offer an opportunity to reduce greenhouse gas emissions and other environmental impacts and can incorporate processes that avoid the use of toxic isocyanates typically used in conventional polyurethane production processes. In this study, the greenhouse gas emissions, fossil energy, and water consumption of biobased polyurethane and biobased non-isocyanate polyurethane were evaluated via life-cycle analysis using the R&D Greenhouse Gases, Regulated Emissions, and Energy Use in Technologies model. Microalgae-based polyurethane foam was found to achieve greenhouse gas emission reductions of up to 79% compared with conventional polyurethane foam production. The greenhouse gas reductions for the non-isocyanate microalgae polyurethane pathway are slightly lower at 58% compared with conventional polyurethane foam. However, it offers additional benefits by reducing toxicity potential compared to the isocyanate polyurethane pathway. The analysis also included a biorefinery-level analysis to evaluate the impact of incorporating polyurethane production into fuel-processing microalgae biorefineries. The sensitivity analyses conducted in this study reveal that improved algae cultivation strategies can lead to decreases of up to 127% and 80% in GHG emissions from the baseline process of Bio-PU and Bio-NIPU, respectively. Likewise, implementation of renewable electricity can result in up to 128% and 74% lower GHG emissions compared to the baseline production of Bio-PU and Bio-NIPU, respectively. Finally, the analysis evaluated different coproduct handling methods including displacement and allocation (based on mass, energy, and market-value). The results suggest that it is important to consider both the displacement and allocation methods as these led to significant differences in the environmental impacts.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Materials Data on PuOF by Materials Project

PuOF crystallizes in the tetragonal P4/nmm space group. The structure is three-dimensional. Pu3+ is bonded in a body-centered cubic geometry to four equivalent O2- and four equivalent F1- atoms. All Pu–O bond lengths are 2.32 Å. All Pu–F bond lengths are 2.59 Å. O2- is bonded to four equivalent Pu3+ atoms to form distorted OPu4 tetrahedra that share corners with four equivalent OPu4 tetrahedra, corners with twelve equivalent FPu4 tetrahedra, edges with two equivalent FPu4 tetrahedra, and edges with four equivalent OPu4 tetrahedra. F1- is bonded to four equivalent Pu3+ atoms to form distorted FPu4 tetrahedra that share corners with four equivalent FPu4 tetrahedra, corners with twelve equivalent OPu4 tetrahedra, edges with two equivalent OPu4 tetrahedra, and edges with four equivalent FPu4 tetrahedra.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Materials Data on Pu2SO2 by Materials Project

Pu2O2S crystallizes in the trigonal P-3m1 space group. The structure is three-dimensional. Pu3+ is bonded in a 7-coordinate geometry to three equivalent S2- and four equivalent O2- atoms. All Pu–S bond lengths are 2.89 Å. There are three shorter (2.32 Å) and one longer (2.35 Å) Pu–O bond lengths. S2- is bonded to six equivalent Pu3+ atoms to form distorted SPu6 octahedra that share corners with twelve equivalent OPu4 tetrahedra, edges with six equivalent SPu6 octahedra, and edges with six equivalent OPu4 tetrahedra. O2- is bonded to four equivalent Pu3+ atoms to form OPu4 tetrahedra that share corners with six equivalent SPu6 octahedra, corners with six equivalent OPu4 tetrahedra, edges with three equivalent SPu6 octahedra, and edges with three equivalent OPu4 tetrahedra. The corner-sharing octahedra tilt angles range from 23–51°.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Materials Data on PuClO by Materials Project

PuOCl is Matlockite structured and crystallizes in the tetragonal P4/nmm space group. The structure is three-dimensional. Pu3+ is bonded in a 9-coordinate geometry to four equivalent O2- and five equivalent Cl1- atoms. All Pu–O bond lengths are 2.32 Å. There are four shorter (3.13 Å) and one longer (3.16 Å) Pu–Cl bond lengths. O2- is bonded to four equivalent Pu3+ atoms to form a mixture of corner and edge-sharing OPu4 tetrahedra. Cl1- is bonded in a 5-coordinate geometry to five equivalent Pu3+ atoms.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Materials Data on PuBrO by Materials Project

PuOBr is Matlockite structured and crystallizes in the tetragonal P4/nmm space group. The structure is two-dimensional and consists of one PuOBr sheet oriented in the (0, 0, 1) direction. Pu3+ is bonded in a 4-coordinate geometry to four equivalent O2- and four equivalent Br1- atoms. All Pu–O bond lengths are 2.30 Å. All Pu–Br bond lengths are 3.23 Å. O2- is bonded to four equivalent Pu3+ atoms to form a mixture of corner and edge-sharing OPu4 tetrahedra. Br1- is bonded in a 4-coordinate geometry to four equivalent Pu3+ atoms.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Materials Data on PuOF by Materials Project

PuOF is half-Heusler structured and crystallizes in the cubic F-43m space group. The structure is three-dimensional. Pu3+ is bonded in a body-centered cubic geometry to four equivalent O2- and four equivalent F1- atoms. All Pu–O bond lengths are 2.43 Å. All Pu–F bond lengths are 2.43 Å. O2- is bonded to four equivalent Pu3+ atoms to form distorted OPu4 tetrahedra that share corners with four equivalent FPu4 tetrahedra, corners with twelve equivalent OPu4 tetrahedra, and edges with six equivalent FPu4 tetrahedra. F1- is bonded to four equivalent Pu3+ atoms to form FPu4 tetrahedra that share corners with four equivalent OPu4 tetrahedra, corners with twelve equivalent FPu4 tetrahedra, and edges with six equivalent OPu4 tetrahedra.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Materials Data on PuIO by Materials Project

PuOI crystallizes in the tetragonal P4/nmm space group. The structure is two-dimensional and consists of one PuOI sheet oriented in the (0, 0, 1) direction. Pu3+ is bonded in a 4-coordinate geometry to four equivalent O2- and four equivalent I1- atoms. All Pu–O bond lengths are 2.32 Å. All Pu–I bond lengths are 3.41 Å. O2- is bonded to four equivalent Pu3+ atoms to form a mixture of edge and corner-sharing OPu4 tetrahedra. I1- is bonded in a 4-coordinate geometry to four equivalent Pu3+ atoms.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Materials Data on PuSbTe by Materials Project

PuSbTe is Matlockite structured and crystallizes in the tetragonal P4/nmm space group. The structure is three-dimensional. Pu5+ is bonded in a 9-coordinate geometry to four equivalent Sb3- and five equivalent Te2- atoms. All Pu–Sb bond lengths are 3.32 Å. There are four shorter (3.15 Å) and one longer (3.24 Å) Pu–Te bond lengths. Sb3- is bonded in a 8-coordinate geometry to four equivalent Pu5+ and four equivalent Sb3- atoms. All Sb–Sb bond lengths are 3.07 Å. Te2- is bonded to five equivalent Pu5+ atoms to form a mixture of distorted edge and corner-sharing TePu5 trigonal bipyramids.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Materials Data on Pu2Te4Cl2O11 by Materials Project

Pu2Te4O11Cl2 crystallizes in the triclinic P-1 space group. The structure is three-dimensional. there are two inequivalent Pu7+ sites. In the first Pu7+ site, Pu7+ is bonded to seven O2- atoms to form a mixture of corner and edge-sharing PuO7 pentagonal bipyramids. There are a spread of Pu–O bond distances ranging from 2.17–2.47 Å. In the second Pu7+ site, Pu7+ is bonded to seven O2- atoms to form a mixture of distorted corner and edge-sharing PuO7 pentagonal bipyramids. There are a spread of Pu–O bond distances ranging from 2.13–2.45 Å. There are four inequivalent Te+2.50+ sites. In the first Te+2.50+ site, Te+2.50+ is bonded in a distorted rectangular see-saw-like geometry to four O2- atoms. There are a spread of Te–O bond distances ranging from 1.87–2.13 Å. In the second Te+2.50+ site, Te+2.50+ is bonded in a 4-coordinate geometry to three O2- and one Cl1- atom. There are a spread of Te–O bond distances ranging from 1.87–1.96 Å. The Te–Cl bond length is 2.79 Å. In the third Te+2.50+ site, Te+2.50+ is bonded in a 3-coordinate geometry to three O2- and one Cl1- atom. There are a spread of Te–O bond distances ranging from 1.87–1.97 Å. The Te–Cl bond length is 2.96 Å. In the fourth Te+2.50+ site, Te+2.50+ is bonded in a 3-coordinate geometry to three O2- and one Cl1- atom. There is two shorter (1.89 Å) and one longer (1.91 Å) Te–O bond length. The Te–Cl bond length is 3.15 Å. There are eleven inequivalent O2- sites. In the first O2- site, O2- is bonded in a distorted trigonal planar geometry to two equivalent Pu7+ and one Te+2.50+ atom. In the second O2- site, O2- is bonded in a distorted bent 150 degrees geometry to one Pu7+ and one Te+2.50+ atom. In the third O2- site, O2- is bonded in a bent 150 degrees geometry to one Pu7+ and one Te+2.50+ atom. In the fourth O2- site, O2- is bonded in a linear geometry to one Pu7+ and one Te+2.50+ atom. In the fifth O2- site, O2- is bonded in a linear geometry to one Pu7+ and one Te+2.50+ atom. In the sixth O2- site, O2- is bonded in a linear geometry to one Pu7+ and one Te+2.50+ atom. In the seventh O2- site, O2- is bonded in a distorted trigonal planar geometry to two Pu7+ and one Te+2.50+ atom. In the eighth O2- site, O2- is bonded in a 3-coordinate geometry to one Pu7+ and two Te+2.50+ atoms. In the ninth O2- site, O2- is bonded in a linear geometry to one Pu7+ and one Te+2.50+ atom. In the tenth O2- site, O2- is bonded in a 3-coordinate geometry to one Pu7+ and two Te+2.50+ atoms. In the eleventh O2- site, O2- is bonded in a distorted trigonal planar geometry to two equivalent Pu7+ and one Te+2.50+ atom. There are two inequivalent Cl1- sites. In the first Cl1- site, Cl1- is bonded in a distorted single-bond geometry to one Te+2.50+ atom. In the second Cl1- site, Cl1- is bonded in a 2-coordinate geometry to two Te+2.50+ atoms.

36 MATERIALS SCIENCE↗