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At least 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Validation of MFUEL Metal Fuel Performance Models of SAS4A/SASSYS-1

The fuel characterization models of SAS4A/SASSYS-1 (SAS) have recently been extended to include a new U-Pu-Zr metal fuel model, MFUEL. MFUEL is equipped with mechanistic physics based models to predict the pre-transient characterization and transient response of metal fuel, with emphasis on fuel melting, cladding failure, and the metal fuel’s impact on core reactivity. The MFUEL model will be available in the full version of SAS4A/SASSYS-1 5.7, which is scheduled to be released in June 2023. Fast reactor fuel pins that operated in EBR-II and FFTF with low smear density U-Zr and U-Pu-Zr metal fuels and irradiation resistant ferritic-martensitic cladding showed significant advantages in achieving high burnups and assuring inherent safety characteristics during anticipated transients, design basis events, and beyond design basis events. For safety analysis, a fuel performance model must be able to predict (1) Fuel pin mechanics and compositional and dimensional changes, (2) Clad failure, and (3) Fuel pin thermal resistance. Achieving these high level goals accurately is strongly related to the model performance of individual physical processes taking place within a fuel pin during its lifetime. Metal fuels typically operate above the mid-point temperature of melting during normal, as well as off-normal, conditions. At these elevated temperatures, the availability of thermal activation provides a driving force for various diffusional processes leading to complex phase transformations, micro-structure evolution, significant amounts of fuel swelling, interconnected porosity formation, excessive amounts of fission gas release, and fuel clad chemical interactions. Clad failure in fast reactors primarily occurs as a result of creep rupture augmented by clad wastage formation. The reaction is driven by thermal creep induced dislocation motion, grain boundary cavity nucleation, growth and breakup of grain boundaries. The high level complexity and limited available data requires introducing physics-based modeling approaches to gain extrapolation ability and sensitivity with respect to various conditions. The objective of this report is to perform validation of MFUEL using the experimental data for (1) Normal operation EBR-II fuel behavior, (2) Normal operation PHENIX fuel behavior, (3) HT9 Pressure tube ramp-and-hold creep rupture tests, (4) Whole Pin Furnace (WPF) creep strain, creep rupture and eutectic tests, (5) Fuel Behavior Test Apparatus (FBTA) eutectic tests, and (6) TREAT M5-7 OverPower tests up to clad failure. Section-2 includes a brief description of the MFUEL models. A detailed description of the MFUEL physics-based, semi-empirical models will be presented in the SAS V 5.7 theory manual. Section-3, 4, and 5 describes the validation effort for the pre-transient irradiation, furnace transients, and TREAT M-Series transients, respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Linear and Non-linear Optical Characterization of Glass and Sapphire for Optical Instrumentation of Advanced Fission Reactors

The linear and nonlinear optical properties of glass and sapphire are investigated to determine the material’s suitability to support optical instrumentation for advanced fission reactors. A post-irradiation examination experiment consisting of a Z-scan, measuring nonlinear optical absorption and refraction, and a spectrometer, measuring linear optical absorption was developed and used. Radiation-induced negative nonlinear optical absorption was observed for the first time in fused silica, sapphire, and borosilicate glass under both types of thermal annealing conditions and the respective effects of thermal annealing are reported.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Observation of new neutron-rich isotopes in the vicinity of Zr 110

Neutron-rich radioactive isotope (RI) beams in the vicinity of 110 Zr were produced using the in-flight fission of a 345-MeV/nucleon 238 U beam at the RIKEN Radioactive Isotope Beam Factory. Here, the RI beams were separated in flight by the large-acceptance two-stage fragment separator BigRIPS. Isotopes were clearly identified in a low-background particle-identification plot, which was obtained from the measurements of time of flight, magnetic rigidity, and energy loss using beam-line detectors placed in the BigRIPS separator and the ZeroDegree spectrometer. Nine new isotopes, 101 Br, 102 Kr, 105,106 Rb, 108 Sr, 110,111 Y, 114 Zr, and 117 Nb, are reported for the first time, with their measured production cross sections compared with the extrapolations of lighter isotopes down to the femtobarn regime.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1976

Application of gaseous core reactors for transmutation of nuclear waste

An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

Schnitzler, B. G.↗
NASA NTRS1980

Spectra from nuclear-excited plasmas

The paper discusses the spectra taken from He-3(n,p)H-3 nuclear-induced plasmas under high thermal neutron flux, lasing conditions. Also, initial spectra are presented for U-235F6 generated plasmas. From an evaluation of these spectra, important atomic and molecular processes that occur in the plasma can be inferred. The spectra presented are the first to be generated by He-3 and U-235F6 nuclear reactions under high neutron flux, lasing conditions. The U-235(n,ff)FF reaction, which liberates 165 MeV of fission-fragment kinetic energy, creates plasmas that are of great interest, since at sufficiently high densities of U-235F6 the gas becomes self-critical; thus, there is no need for an external driving reactor (source of neutrons). The spectra from mixtures of He-3 and Ar, Xe, Kr, Ne, Cl2, F2 and N2 indicate little difference between high-pressure nuclear-induced plasmas and high-pressure electrically pulsed afterglow plasmas for noble-gas systems

De Young, R. J.↗
DOE OSTI2022

Monitoring Noble Gases (Xe and Kr) and Aerosols (Cs and Rb) in a Molten Salt Reactor Surrogate Off-Gas Stream Using Laser-Induced Breakdown Spectroscopy (LIBS)

In this study with surrogate materials we show that laser-induced breakdown spectroscopy (LIBS) is a robust tool with promising capability toward monitoring gaseous (Xe and Kr) and aerosol (Cs and Rb) species in an off-gas stream from a molten salt reactor (MSR). MSRs will continually evolve fission products into the cover gas flowing across the reactor headspace. The cover gas entrains Xe and Kr gases, along with aerosol particles, before passing into an off-gas treatment system. Univariate models of Xe and Kr peaks showed a strong correlation to concentration indicated by their coefficients of determination of 0.983 and 0.997, respectively. Multivariate models were built for all four analytes using partial least squares regression coupled with preprocessing steps including normalization, trimming, and/or genetic algorithm derived filters. The models were evaluated by predicting the concentrations of the analytes in four validation samples, in which all calibration models were successfully validated at a confidence interval of 99.9%. Finally, pressure controllers were used to regulate the mass flow rate of Kr flowing into the measurement cell in sinusoidal and stepwise waveforms to test the real-time monitoring capabilities of the regression models. Both univariate and partial least squares Kr models were able to successfully quantify the gas concentration in the real-time evaluation. The root mean squared error of prediction (RMSEP) values for these real-time tests were calculated to be 0.051, 0.060, and 0.121 mol% demonstrating the measurement systems’ capability to perform online monitoring with acceptable accuracy.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Improved irradiation resistance of accident-tolerant high-strength FeCrAl alloys with heterogeneous structures

In this work, post–neutron irradiation examination is performed on advanced accident-tolerant fuel (ATF) cladding iron-chromium-aluminum (FeCrAl) alloys with ~10–13at. % Cr, ~10–12 at. % Al, ~1 at. % Mo, and minor alloying elements including Y irradiated to a damage level of 7 displacements per atom (dpa) at irradiation temperatures of 267–282 °C. A compositional dependency of the Cr and Al content is observed on the ratio of sessile and glissile dislocation loops, where the density of a$\langle$100$\rangle$ type loops is somewhat higher than the a/2$\langle$111$\rangle$ type loops. The α' precipitate number density is inversely correlated to the starting Cr concentration of the alloys of interest. The irradiation to a higher dose of 7 dpa results in a higher density of dislocation loops and α' precipitates for the same alloys at a lower irradiation dose, such as 1.8 dpa. In this work, the effect of α' precipitates on the dislocation loop density is discussed, and the presence of α' appears to inhibit the nucleation of loops. Compared with first-generation FeCrAl alloys, these advanced alloys with heterogeneous structure exhibit a lower Cr concentration in α' precipitation at the same dose level; they act as weaker obstacles deviating from the primary hardening contribution from the mature α'. Hence, the overall irradiation-induced hardening decreases; our alloys show improved radiation resistance because of their stronger sink strengths. The results presented in this paper could provide insights for the design and optimization of ATF cladding materials for future fission and space applications.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Microstructure and diffusion behavior of uranium fuel with minor additives

Fission product lanthanides in metallic fuels are known to cause adverse fuel-cladding chemical interaction (FCCI). Tin (Sn) and palladium (Pd) are being explored as the potential additives to reduce or mitigate lanthanide-induced FCCI by forming stable Sn-Ln and Pd-Ln compounds. The current study is an investigation of the fuel alloys, U–4Sn, U–4Sn-4Ln, U–4Pd, and U–4Pd-4Ln (wt. %), and their diffusion behaviors with Fe. Microstructural analysis was performed using scanning electron microscopy (SEM). Further, the binary phases in the UPd and USn alloys are identified as UPd 3 and suspected U 2 Sn, where U 2 Sn is unknown in the literature. The binary phases in the UPdLn alloy are identified as PdLn and Pd-rich Pd-Ln (Pd 4 Ln 3 and Pd 3 Ln 2 ), and in the USnLn alloy is SnLn. Diffusion of Fe with these binary phases is insignificant compared to the diffusion with U. The Ln-induced FCCI can be mitigated by the additives.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Mn-doped oxide nuclear fuel

A nuclear fuel includes uranium(IV) oxide (UO2) and manganese (Mn) as a dopant. The Mn dopant may be present in the fuel in an amount up to the solubility limit for Mn under a given set of conditions, for example, about 0.01 wt % to about 1 wt %. The nuclear fuel is substantially free of aluminum (Al). The nuclear fuel exhibits enhanced grain size development during sintering temperatures as low at 1400 K due to an increase in uranium sub-lattice vacancies induced by dissolution of the Mn dopant at interstitial defect sites. The Mn-doped nuclear fuel exhibits improved grain sizes at lower temperatures compared to Cr-, Al-, and undoped UO2, and therefore desirably exhibits lower fission gas release and higher plasticity, reducing the chances of fuel rod failure.

Cooper, Michael William Donald↗
DOE OSTI2024

Multipole responses in fissioning nuclei and their uncertainties

Electromagnetic multipole responses are key inputs to model the structure, decay, and reactions of atomic nuclei. With the introduction of the finite amplitude method (FAM), large-scale calculations of the nuclear linear response in heavy deformed nuclei have become possible. This work provides a detailed study of multipole responses in actinide nuclei with Skyrme energy density functionals. We quantify both systematic and statistical uncertainties induced by the functional parametrization in FAM calculations. We also extend the FAM formalism to perform blocking calculations with the equal filling approximation for odd-mass and odd-odd nuclei, and analyze the impact of blocking configurations on the response. In conclusion, by examining the entire plutonium isotopic chain from the proton to the neutron drip line, we find a large variability of the response with the neutron number and study how it correlates with the deformation of the nuclear ground state.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Entanglement entropy, single-particle occupation probabilities, and short-range correlations

For quantum many-body systems with short-range correlations (SRCs), the intimate relationship between their magnitude, the behavior of the single-particle occupation probabilities at momenta larger than the Fermi momentum, and the entanglement entropy is a new qualitative aspect not studied and exploited yet. A large body of recent condensed matter studies indicates that the time evolution of the entanglement entropy describes the nonequilibrium dynamics of isolated and strongly interacting many-body systems, in a manner similar to the Boltzmann entropy, which is strictly defined for dilute and weakly interacting many-body systems. Both theoretical and experimental studies in nuclei and cold atomic gases have shown that the fermion momentum distribution has a generic behavior n(k)=C/k 4 at momenta larger than the Fermi momentum, due to the presence of SRCs, with approximately 20% of the particles having momenta larger than the Fermi momentum. Further, the presence of the long momentum tails in the presence of SRCs changes the textbook relation between the single-particle kinetic energy and occupation probabilities, n mf ⁡(k) = 1/{1+ exp ⁡β[ϵ⁡(k)-μ]} for momenta very different form the Fermi momentum, particularly for dynamics processes. SRCs induced high-momentum tails of the single-particle occupation probabilities increase the entanglement entropy of fermionic systems, which in its turn affects the dynamics of many nuclear reactions, such as heavy-ion collisions and nuclear fission.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

A Molten Salt Aerosol Spectroscopy Approach to Online Process Monitoring in Molten Salt Reactors

Recent research and development in molten salt reactors (MSRs) has led to many advanced reactor designs, many of which have demonstrations planned in the next decade. However, a significant challenge that must be overcome before liquid fueled MSRs become viable globally is nuclear material accountancy (NMA) and safeguards. Liquid-fueled MSRs are problematic to safeguards because the nuclear material is in the bulk phase, inventories are constantly changing, and there are potentially many outlet pathways that must be monitored, especially in designs with online chemical processing. At the Idaho National Laboratory, we have developed a process monitoring tool to circulate molten salt through a nebulizer and optical cell, where laser-induced breakdown spectroscopy (LIBS) and ultraviolet-visible (UV-VIS) are applied to track the elemental and oxidation states of the constituents in the salt aerosol. This approach has the potential to provide information on the actinides, fission products, and corrosion in the reactor or process. Preliminary testing in aqueous praseodymium chloride (PrCl3) and neodymium chloride (NdCl3) showed good sensitivity detection limits n developed calibration curves. Additional tests and analysis or ongoing in aqueous solutions while equipment is being scaled up for high temperature molten salt testing. Results for continued testing as well as a status of the molten salt design and anticipated commissioning will be presented.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
NASA NTRS1979

Research on radiation induced laser plasmas

The development of high power nuclear pumped lasers is discussed. The excitation mechanism of continuous wave (CW) HeNe nuclear pumped lasers is studied and a CO2 nuclear pumped laser is used to demonstrate the CW output in the order of watts. The assumption that high power densities are only achievable by volume fission fragment sources is used to identify laser gases which are compatible with UF6 by excited states lifetime measurements. The examination of Xe2, XeF, and KrF under nuclear irradiation to determine if they are good candidates for nuclear-pumped lasers is described.

Schneider, R. T.↗
DOE OSTI2024

Compositionally complex carbide ceramics: A perspective on irradiation damage

Extensive experimental and computational studies have demonstrated outstanding physical and chemical properties of the novel materials of compositionally complex carbides (CCCs), enabling their promising applications in advanced fission and fusion energy systems. This perspective provides a comprehensive overview of radiation damage behavior reported in the literature to understand the fundamental mechanisms related to the impact of multi-principal metal components on phase stability, irradiation-induced defect clusters, irradiation hardening, and thermal conductivity of compositionally complex carbides. Several future research directions are recommended to critically evaluate the feasibility of designing and developing new ceramic materials for extreme environments using the transformative “multi-principal component” concept. Compared to the existing materials for nuclear applications including stainless steels, nickel alloys, ZrC, SiC, and potentially high-entropy alloys, as well as certain other compositionally complex ceramic families. CCCs appear to be more resistant to amorphization, growth of irradiation defect clusters, and void swelling.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Fission gas retention of densely packed uranium carbonitride tristructural-isotropic fuel particles in a 3D printed SiC matrix

The Transformational Challenge Reactor (TCR) fuel form was designed to contain large, densely packed uranium carbonitride (UCN) tristructural-isotropic (TRISO) fuel particles within a 3D printed SiC matrix, increasing the uranium density compared to conventional TRISO fuel forms and offering full geometric freedom for core design. Here, this work summarizes initial low-burnup, high-power irradiation testing of TCR fuel materials, including loose UCN TRISO particles and integral fuel compacts with ~55% TRISO particles by volume, to evaluate fission gas retention. Fission gasses were fully retained in all loose particle tests and in integral compacts irradiated at low (<250 °C) surface temperatures. Initial testing at higher (~700–750 °C) fuel surface temperatures showed fission gas release (FGR) and complete fracture of three compacts, but no FGR was observed in later high temperature tests (~300–750 °C) of both fueled compacts and loose TRISO particles. Calculated thermal stresses in the failed compacts were far less than the measured strength of the SiC matrix and the stresses in some failed compacts were less than those in compacts that did not show FGR. Thermal stress-induced matrix cracks also would not cause complete fracture because the tensile stresses transition to compression in the higher temperature regions. Therefore, fuel failure was likely not caused by thermal stresses and may have been related to leakage currents from the electrical heaters and erratic fuel surface temperatures that were only observed in the test for which failure was observed. In any case, the matrix cracks propagated through the coatings of TRISO particles located in the high-density matrix regions on the peripheries of the compacts, resulting in measurable fission gas release. The discussion focuses on the importance of understanding matrix density distributions and the particle-matrix interface properties to prevent matrix cracks from causing TRISO particle failures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Autoclave grid-to-rod fretting wear evaluation of a candidate cladding coating for accident-tolerant fuel

In pressurized water reactors (PWRs), water flow induced vibrations cause contact and rubbing between the fuel rods and the supporting grid, a phenomenon known as Grid-to-Rod-Fretting (GTRF). GTRF may produce progressive wear damage on the fuel claddings leading to subsequent leakage of radioactive fission products. Various accident-tolerant fuel (ATF) concepts are being developed for higher resistance to the high temperature steam and one approach is to apply a cladding coating. Here, fretting wear behavior of a candidate Cr-coating was investigated using a unique bench-scale autoclave testing rig mimicking the environment in an industrial full-assembly PWR simulator. The contact was under a realistically low load (~0.5 N) lubricated by deionized water at a temperature of 204 °C under a pressure of 20-23 bars. Results demonstrated that the Cr-coating significantly improved the cladding's wear resistance when tested against a commercial ZIRLO grid with or without pre-oxidization. In addition, the Cr-coating also reduced wear on the non-oxidized ZIRLO grid but slightly increased the wear on the pre-oxidized grid.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1989

Ecology and life history of an amoebomastigote, Paratetramitus jugosus, from a microbial mat: new evidence for multiple fission

Five microbial habitats (gypsum crust, gypsum photosynthetic community, Microcoleus mat, Thiocapsa scum, and black mud) were sampled for the presence of the euryhaline, rapidly growing amoebomastigote, Paratetramitus jugosus. Field investigations of microbial mats from Baja California Norte, Mexico, and Salina Bido near Matanzas, Cuba, reveal that P. jugosus is most frequently found in the Thiocapsa layer of microbial mats. Various stages of the life history were studied using phase-contrast, differential-interference, and transmission electron microscopy. Mastigote stages were induced and studied by electron microscopy; mastigotes that actively feed on bacteria bear two or more undulipodia. A three-dimensional drawing of the kinetid ("basal apparatus") based on electron micrographs is presented. Although promitoses were occasionally observed, it is unlikely that they can account for the rapid growth of P. jugosus populations on culture media. Dense, refractile, spherical, and irregular-shaped bodies were seen at all times in all cultures along with small mononucleate (approximately 2-7 micrometers diameter) amoebae. Cytochemical studies employing two different fluorescent stains for DNA (DAPI, mithramycin) verified the presence of DNA in these small bodies. Chromatin-like material seen in electron micrographs within the cytoplasm and blebbing off nuclei were interpreted to the chromatin bodies. Our interpretation, consistent with the data but not proven, is that propagation by multiple fission of released chromatin bodies that become small amoebae may occur in Paratetramitus jugosus. These observations are consistent with descriptions of amoeba propagules in the early literature (Hogue, 1914).

Non-NASA Center↗
DOE OSTI2022

Recent Advances in Radiation-Induced Actinide Redox Chemistry

The actinide series boasts many unique physical and chemical features worthy of both fundamental and applied study. However, the chemical influence of their inherent radiation field is often overlooked, especially as we begin to explore the late actinides in more detail than ever possible before. From the perspective of used nuclear fuel reprocessing, the absorption of ionizing radiation induces the formation of a variety of transient and steady-state excited states, radicals, ions, and molecular degradation products, many of which are highly redox active and can lead to significant changes in a reprocessing solvent system’s physical and chemical properties. For example, radiolysis of the actinides can drive steady-state redox distributions and the formation of non-traditional oxidation states which can complicate their separation and recovery from fission products. This scenario is further exacerbated when complexation is taken into account. Consequently, a molecular-level understanding of radiation effects on the actinides over multiple time, distance, and material domains is essential for supporting innovation in used nuclear fuel reprocessing technologies. Attaining this knowledge necessitates a firm grasp of actinide radiation chemistry to develop predictive, mechanistic, multiscale models to support engineering efforts. Presented here are several recent studies that highlight recent advances in actinide radiation chemistry, in particular, the effect of actinide complexation on ligand reactivity towards radiation-induced transients.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗