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UO2(NH4)3(CO2)4NH2OH2O crystallizes in the monoclinic P2_1/c space group. The structure is zero-dimensional and consists of twelve ammonium molecules; sixteen formic acid molecules; four hydroxylamine molecules; four pitchblend, uranium(iv) oxide, uranium(iv) dioxide, uranium dioxide molecules; and four water molecules.
UO2(NO3)2(H2O)6 crystallizes in the orthorhombic Cmc2_1 space group. The structure is zero-dimensional and consists of eight nitric acid molecules; four pitchblend, uranium(iv) oxide, uranium(iv) dioxide, uranium dioxide molecules; and twenty-four water molecules.
UO2(NO)2(Br)4 crystallizes in the triclinic P-1 space group. The structure is zero-dimensional and consists of one 1,3,2,4-dioxadiazetidine molecule; four hydrobromic acid molecules; and one pitchblend, uranium(iv) oxide, uranium(iv) dioxide, uranium dioxide molecule.
UO2(CO2)4 crystallizes in the orthorhombic Ccce space group. The structure is zero-dimensional and consists of sixteen formic acid molecules and four pitchblend, uranium(iv) oxide, uranium(iv) dioxide, uranium dioxide molecules.
Cs3(UO2)2(PO4)O2 crystallizes in the monoclinic C2/c space group. The structure is three-dimensional. there are two inequivalent Cs1+ sites. In the first Cs1+ site, Cs1+ is bonded in a 5-coordinate geometry to eight O2- atoms. There are a spread of Cs–O bond distances ranging from 3.12–3.60 Å. In the second Cs1+ site, Cs1+ is bonded in a 6-coordinate geometry to six O2- atoms. There are a spread of Cs–O bond distances ranging from 3.02–3.18 Å. There are two inequivalent U6+ sites. In the first U6+ site, U6+ is bonded to six O2- atoms to form UO6 octahedra that share corners with two equivalent UO6 octahedra and corners with two equivalent PO4 tetrahedra. The corner-sharing octahedral tilt angles are 43°. There are a spread of U–O bond distances ranging from 1.88–2.42 Å. In the second U6+ site, U6+ is bonded to six O2- atoms to form UO6 octahedra that share corners with two equivalent UO6 octahedra and corners with two equivalent PO4 tetrahedra. The corner-sharing octahedral tilt angles are 43°. There are a spread of U–O bond distances ranging from 1.89–2.30 Å. P5+ is bonded to four O2- atoms to form PO4 tetrahedra that share corners with four UO6 octahedra. The corner-sharing octahedra tilt angles range from 34–53°. There is two shorter (1.55 Å) and two longer (1.57 Å) P–O bond length. There are five inequivalent O2- sites. In the first O2- site, O2- is bonded in a 1-coordinate geometry to two Cs1+, one U6+, and one P5+ atom. In the second O2- site, O2- is bonded in a distorted single-bond geometry to three Cs1+ and one U6+ atom. In the third O2- site, O2- is bonded in a distorted bent 150 degrees geometry to one Cs1+, one U6+, and one P5+ atom. In the fourth O2- site, O2- is bonded in a 2-coordinate geometry to two equivalent Cs1+ and two U6+ atoms. In the fifth O2- site, O2- is bonded in a distorted single-bond geometry to three Cs1+ and one U6+ atom.
(UO2)2(BO2)9 crystallizes in the triclinic P1 space group. The structure is two-dimensional and consists of two pitchblend, uranium(iv) oxide, uranium(iv) dioxide, uranium dioxide molecules and one BO2 sheet oriented in the (0, 0, 1) direction. In the BO2 sheet, there are nine inequivalent B sites. In the first B site, B is bonded to four O atoms to form corner-sharing BO4 tetrahedra. There are a spread of B–O bond distances ranging from 1.41–1.50 Å. In the second B site, B is bonded to four O atoms to form corner-sharing BO4 tetrahedra. There are a spread of B–O bond distances ranging from 1.45–1.49 Å. In the third B site, B is bonded in a trigonal planar geometry to three O atoms. There is one shorter (1.37 Å) and two longer (1.38 Å) B–O bond length. In the fourth B site, B is bonded in a trigonal planar geometry to three O atoms. There is one shorter (1.37 Å) and two longer (1.38 Å) B–O bond length. In the fifth B site, B is bonded to four O atoms to form corner-sharing BO4 tetrahedra. There are a spread of B–O bond distances ranging from 1.45–1.49 Å. In the sixth B site, B is bonded to four O atoms to form corner-sharing BO4 tetrahedra. There are a spread of B–O bond distances ranging from 1.44–1.53 Å. In the seventh B site, B is bonded to four O atoms to form corner-sharing BO4 tetrahedra. There are a spread of B–O bond distances ranging from 1.45–1.48 Å. In the eighth B site, B is bonded to four O atoms to form corner-sharing BO4 tetrahedra. There are a spread of B–O bond distances ranging from 1.45–1.47 Å. In the ninth B site, B is bonded in a trigonal planar geometry to three O atoms. There are a spread of B–O bond distances ranging from 1.34–1.39 Å. There are eighteen inequivalent O sites. In the first O site, O is bonded in a bent 120 degrees geometry to two B atoms. In the second O site, O is bonded in a distorted bent 120 degrees geometry to two B atoms. In the third O site, O is bonded in a linear geometry to two B atoms. In the fourth O site, O is bonded in a bent 120 degrees geometry to two B atoms. In the fifth O site, O is bonded in a single-bond geometry to one B atom. In the sixth O site, O is bonded in a single-bond geometry to one B atom. In the seventh O site, O is bonded in a distorted bent 120 degrees geometry to two B atoms. In the eighth O site, O is bonded in a distorted bent 120 degrees geometry to two B atoms. In the ninth O site, O is bonded in a distorted bent 120 degrees geometry to two B atoms. In the tenth O site, O is bonded in a distorted bent 120 degrees geometry to two B atoms. In the eleventh O site, O is bonded in a distorted bent 120 degrees geometry to two B atoms. In the twelfth O site, O is bonded in a distorted bent 120 degrees geometry to two B atoms. In the thirteenth O site, O is bonded in a single-bond geometry to one B atom. In the fourteenth O site, O is bonded in a bent 120 degrees geometry to two B atoms. In the fifteenth O site, O is bonded in a distorted bent 120 degrees geometry to two B atoms. In the sixteenth O site, O is bonded in a distorted bent 120 degrees geometry to two B atoms. In the seventeenth O site, O is bonded in a distorted bent 120 degrees geometry to two B atoms. In the eighteenth O site, O is bonded in a distorted water-like geometry to two B atoms.
UO2(OH)4 crystallizes in the orthorhombic Pnma space group. The structure is three-dimensional. U is bonded to eight O atoms to form distorted edge-sharing UO8 hexagonal bipyramids. There are a spread of U–O bond distances ranging from 1.80–2.44 Å. There are two inequivalent H sites. In the first H site, H is bonded in a single-bond geometry to two O atoms. There is one shorter (1.00 Å) and one longer (1.76 Å) H–O bond length. In the second H site, H is bonded in a single-bond geometry to one O atom. The H–O bond length is 0.99 Å. There are four inequivalent O sites. In the first O site, O is bonded in a single-bond geometry to one U atom. In the second O site, O is bonded in a single-bond geometry to one U atom. In the third O site, O is bonded in a 4-coordinate geometry to two equivalent U, one H, and one O atom. The O–O bond length is 1.45 Å. In the fourth O site, O is bonded in a distorted water-like geometry to one U and two H atoms.
UO2(N2)2(CO3)3 crystallizes in the monoclinic C2/c space group. The structure is zero-dimensional and consists of sixteen ammonia molecules; twelve carbonic acid molecules; and four pitchblend, uranium(iv) oxide, uranium(iv) dioxide, uranium dioxide molecules.
UO2(CH3)3C6NH16(CO2)3 crystallizes in the orthorhombic Pna2_1 space group. The structure is zero-dimensional and consists of twelve formic acid molecules; twelve methane molecules; four pitchblend, uranium(iv) oxide, uranium(iv) dioxide, uranium dioxide molecules; and four triethylammonium molecules.
Uranyl oxy-hydroxy-hydrate minerals are common alteration products of uraninite (UO2+x) which is chemically and structurally analogous to uranium dioxide nuclear fuel. Therefore, structural and spectroscopic investigations of these alteration minerals and their analogous anthropogenic counterparts can provide insight into the environmental behavior of nuclear fuel cycle materials. Previously, we compiled available vibrational spectroscopic data for uranyl minerals in the Compendium of Uranium Raman and Infrared Experimental Spectra (CURIES) and found that only 37% of known uranyl oxy-hydroxy-hydrate minerals had spectra readily available in the literature and existing databases for inclusion therein. Furthermore, no available infrared spectra for this mineral group were included in CURIES. To expand our understanding of the spectroscopic features of uranyl oxy-hydroxy-hydrates and the structural origins thereof, we collected, and now include in CURIES, Raman and infrared spectra for an additional 12 uranyl hydroxide phases. To better understand the impact of structural and compositional variations of these phases on their spectroscopic features, we compare Raman spectra of different anion sheet topological groups and of analogous phases hosting different counter cations. We identify spectroscopic variations related to differences in equatorial bonding and structural changes as a result of cation substitution. We also prepare a uranyl hydroxide phase via hydrolysis of uranyl fluoride (UO2F2) as an analog of hydrolysis reactions that occur in nuclear fuel cycle materials; and we find that the alteration product of UO2F2, despite chemical and structural similarities to uranyl oxy-hydroxy-hydrate minerals, is readily distinguishable from related mineral phases using Raman spectroscopy. In this work, we provide new insights into the structural origins of spectroscopic features in uranyl oxy-hydroxy-hydrate minerals, improve the average Raman spectrum for this group of minerals, and thereby improve capabilities for identifying these mineral species and related anthropogenic phases using Raman spectroscopy.
After a tsunami triggered plant black out at the Fukushima Daiichi site, followed by hydrogen explosions, the US Department of Energy partnered with fuel vendors to study safer alternatives to the current UO2 – zirconium alloy fuel system. This accident tolerant fuel (ATF) alternative should better tolerate a loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General Electric, Oak Ridge National Laboratory, and their partners are proposing to replace zirconium alloy cladding in current commercial light water power reactors with an iron-chromium-aluminum (FeCrAl) cladding such as APMT, C26M or FeCrAl ODS, which has been named IronClad by GE. Extensive testing and evaluation has been conducted to determine the suitability of FeCrAl for the entire fuel cycle from rod fabrication to used fuel disposition. Most of the environmental characterization was conducted to evaluate the behavior of IronClad under normal operation conditions and under severe accident conditions. Results show that FeCrAl has excellent corrosion resistance under normal operation conditions and FeCrAl is several orders of magnitude more resistant than zirconium alloys to degradation by superheated steam under accident conditions, generating less heat of oxidation and lower amount of combustible hydrogen gas. Higher parasitic neutron absorption and potential tritium release into the coolant effects can be minimized by design and operational measures. IronClad material can be manufactured using traditional melting and forging (C26M) or powder metallurgy processes (APMT & FeCrAl ODS). Tubes can be made to full length of 5 m with a wall thickness of 0.4 mm by pilgering or drawing. The fuel rods can be readily hermetically sealed in an industrial production setting by welding end caps to the thin wall tubing using either TIG or resistance welding. The implementation of FeCrAl IronClad cladding is a near term solution to enhance the safety of the current fleet of commercial light water power reactors.
The reprocessing of used nuclear fuel would release volatile radionuclides into the off-gas of a processing plant, including 3 H, 14 C, 85 Kr, and 129 I. One potential simplification to the management of the off-gas streams could be achieved through an efficient tritium pretreatment (TPT) step in which the UO 2 fuel is oxidized either by air or by NO2 before dissolution. Oxidation of the UO2 fuel matrix results in the release of tritium contained within. Upfront removal of tritium from the fuel in a pretreatment step, followed by the abatement of tritium, can minimize or eliminate the distribution of tritium throughout the plant, also decreasing or eliminating the need for tritium capture on multiple off-gas streams. The use of NO 2 as the oxidant in an advanced tritium pretreatment (ATPT) allow the oxidation to be performed at lower temperatures and may result in the quantitative release of iodine from the fuel.
Post-irradiation examination (PIE) and elevated-temperature safety testing are being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s second irradiation experiment (AGR-2). Details on this irradiation experiment have been previously reported. The AGR-2 PIE effort builds upon the understanding acquired throughout the AGR-1 PIE campaign (Demkowicz et al. 2015) and is establishing a database for the different AGR-2 fuel designs. The AGR-2 irradiation experiment included TRISO fuel particles coated at BWX Technologies (BWXT) with an engineering-scale coater, which had a coating chamber 150 mm in diameter. Two coating batches were tested in the AGR-2 irradiation experiment. Batch G73H-10-93085B had uranium dioxide (UO2) kernels with an average diameter of 508 µm. Batch G83J-14-93073A had UCO kernels with an average diameter of 427 µm; in this kernel design, some of the uranium oxide is converted to uranium carbide during fabrication to provide a getter for oxygen liberated during fission and to limit CO production. Fabrication and property data for the AGR-2 coating batches have been compiled and compared to AGR-1. The AGR-2 TRISO coatings were most like the AGR-1 Variant 3 TRISO deposited in the Oak Ridge National Laboratory (ORNL) laboratory-scale coater, which had a coating chamber 50 mm in diameter (Hunn and Lowden 2006). In both cases, the hydrogen and methyltrichlorosilane coating gas mixture that was used to deposit the SiC was diluted with argon to produce a finer grained, more equiaxed SiC microstructure. In addition to the fact that AGR-1 fuel had smaller UCO kernels that were 350 µm in diameter, other notable differences in the TRISO particle properties included (1) the pyrocarbon anisotropy, which was slightly higher in the particles coated in the engineering-scale coater, and (2) the exposed kernel defect fraction, which was higher for AGR-2 fuel due to the detected presence of particles with impact damage that was introduced during TRISO particle handling.
Post-irradiation examination (PIE) and elevated-temperature safety testing are being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s second irradiation experiment (AGR-2). Details on this irradiation experiment have been previously reported. The AGR-2 PIE effort builds upon the understanding acquired throughout the AGR-1 PIE campaign and is establishing a database for the different AGR-2 fuel designs. The AGR-2 irradiation experiment included TRISO fuel particles coated at BWX Technologies (BWXT) with an engineering-scale coater, which had a coating chamber 150 mm in diameter. Two coating batches were tested in the AGR-2 irradiation experiment. Batch G73H-10-93085B had uranium dioxide (UO2) kernels with an average diameter of 508 µm. Batch G83J-14-93073A had UCO kernels with an average diameter of 427 µm; in this kernel design, some of the uranium oxide is converted to uranium carbide during fabrication to provide a getter for oxygen liberated during fission and to limit CO production. Fabrication and property data for the AGR-2 coating batches have been compiled and compared to AGR-1. The AGR-2 TRISO coatings were most like the AGR-1 Variant 3 TRISO deposited in the Oak Ridge National Laboratory (ORNL) laboratory-scale coater, which had a coating chamber 50 mm in diameter (Hunn and Lowden 2006). In both cases, the hydrogen and methyltrichlorosilane coating gas mixture that was used to deposit the SiC was diluted with argon to produce a finer grained, more equiaxed SiC microstructure. In addition to the fact that AGR-1 fuel had smaller UCO kernels that were 350 µm in diameter, other notable differences in the TRISO particle properties included (1) the pyrocarbon anisotropy, which was slightly higher in the particles coated in the engineering-scale coater, and (2) the exposed kernel defect fraction, which was higher for AGR-2 fuel due to the detected presence of particles with impact damage that was introduced during TRISO particle handling.
In support of the Accident Tolerant Fuels program, experiment devices are being developed to enable transient testing of pressurized water reactor (PWR) type fuel specimens. In fiscal year 2019 several fueled tests were conducted in the Transient Reactor Test Facility (TREAT) using the Separate Effects Test Holder (SETH). Five different SETH capsules (SETH A-E) were assembled, each holding a 10-pellet rodlet with 4.9% enriched, commercially-produced fresh UO2 pellets in zirconium alloy cladding with pellet dimensions having typical PWR radial dimensions and a total pellet stack length measuring 10.16 cm. The rodlet’s length allows the specimen to be axially positioned in the center of the core with minimal flux variation over the height of the rodlet. The fuel motion monitoring system (FMMS) is located on the north side of the TREAT reactor. Having direct line-of-sight of the core enables the FMMS to collect spatially-resolved data for fast neutrons generated due to fission occurring in experiment specimens. The FMMS contains 360 channels that make up a 10 (horizontal) by 36 (vertical) array of viewing slots; for the SETH A-E test series 96 proton-recoil scintillator (PRS) detectors were installed in the system. The fast neutrons are detected in the PRS detectors from light produced by the recoil of protons interacting with ZnS(Ag) grains in each detector's scintillator button. This report focuses on the FMMS data captured in the SETH A-E experiments, with multiple transients varying in reactivity addition and length. The first three experiments had a total reactor energy of 101 MJ each and were critical in understanding the FMMS field of view, lead shielding, and other TREAT data acquisition systems. SETH-B2 and C had a reactivity addition of almost double the previous transients and served as a demonstration of the detector array behavior under increased reactor energy. The final two SETH experiments (D and E) were aimed at reaching cladding melting temperatures with a reactor energy over 500 MJ for each test. In preparation for the higher energy transients, the FMMS lead shielding was reduced to 0.5 inches from the 6 inches used in SETH A-C. This was expected to saturate the detectors at the peak of the transient in exchange for higher sensitivity at lower energies after the peak, when the motion was expected. SETH-D was clipped at approximately 3 seconds, which was too early for the FMMS to capture any rodlet motion. SETH-E had the same reactivity addition as SETH-D but the transient time was extended to 18 seconds. The FMMS captured the downward motion of the rodlet in SETH-E as the count rate significantly changed for the detectors viewing the specimen. Snapshots taken throughout the transient helped identify the approximate time at which the motion initiated. Further analysis of the individual detector data captured on a per-millisecond time basis resulted in a determination that the approximate time for the onset of the downward motion was 2.84 ± 0.02 seconds. Neutron radiographs confirmed that downward motion occurred for SETH-D and SETH-E. Motion in SETH-D could also have been captured by the FMMS if the transient time had been longer for SETH-D. Ultimately, these final two tests served to demonstrate the FMMS's ability to track fuel motion during transient experiments at TREAT.
The Transient Reactor Test (TREAT) facility is a graphite based transient test reactor at the Idaho National Laboratory (INL). TREAT started operation in 1959 and performed thousands of transient irradiations for the primary purpose of nuclear fuel safety research. TREAT operations were suspended in 1994, and then, several years later, resumed reactor operations in 2017 to support development of advanced nuclear fuels. TREAT’s more recent historic focus had been for transients with relevance to sodium cooled fast spectrum reactors. TREAT’s first year of modern experiment work performed in 2018, however, saw the demonstration of new kinds of transients focused on light water reactor (LWR) needs including reactivity initiated accidents (RIA) and loss of coolant accidents (LOCA). The successful outcome of these efforts confirmed TREAT’s viability as a transient neutron source for LWR safety research. Since TREAT’s core is air cooled and provides no “built-in” thermal hydraulic boundary conditions for tests specimens, efforts have been underway to establish water-environment capabilities for testing specimens. Small static water capsules have been established to support single rod RIA testing where transient heating is brief. Similarly, deployment of enhanced water capsules with blowdown capability is nearing completion to support single rod LOCA testing. With careful experiment design these capsules can provide representative temperature response histories and water phase transitions for some experiment parameters. The capsule approach is appropriate for cost-effective early phase phenomena identification studies on novel fuels, or when the important test variables are thought to be fairly well understood so that they can be precisely targeted (such as they are for current fleet fuels with moderate burnup UO2 in uncoated zirconium alloy cladding). Due to lack of forced convection, however, the capsule approach cannot achieve representative temperature histories in all regions of the test fuel simultaneously (e.g. cladding and fuel centerline temperatures) while providing representative timing in water phase transitions. Nor can the capsule approach provide appropriate hydraulic conditions for multi-rod bundle testing. For these reasons, design and development has been undertaken for the TREAT Water-Environment Recirculating Loop (TWERL).
This report contains the results of non-destructive examinations of Accident Tolerant Fuels 2 (ATF-2) rodlets irradiated in the Advanced Test Reactor (ATR) Loop 2A. The experiment is part of the U.S. Department of Energy Nuclear Technology Research and Development (NTRD) program’s Advanced Fuels Campaign (AFC). The rodlets were composed of UO2 pellets and Zr-4 cladding. The data have been collected to provide baseline PIE data to support future testing in TREAT and to have performance data against which the data of the ATF concepts can be compared directly. The analyses performed included: visual examinations, axial gamma scanning, neutron radiography and profilometry. All the data collected showed a performance consistent with expectations for this fuel system at low burnup.
FeCrAl-UO2 test capsules were fabricated at Oak Ridge National Laboratory (ORNL) and irradiated at the Advanced Test Reactor. Following irradiation, samples were sectioned from the irradiated rod and oxidation kinetics were evaluated to access the candidate cladding high-temperature oxidation performance following irradiation. The irradiation was performed at approximately 400°C to a burnup of 10 GWd/MT. The high-temperature oxidation tests were conducted in the ORNL Severe Accident Test Station at 1200 and 1300°C. Weight measurements were taken before and after oxidation testing. Cross-sections of the cladding were metallographically mounted and optical microscopy was performed. Measurements of the oxidation layer before and after high-temperature testing were collected. The results indicate the irradiated FeCrAl C35M alloy provided good thermal stability up to 1200°C.